ML20134Q136

Ack Receipt of Faxes from Licensee Re Unit 2 Cycle 6 Rod Control Assembly Evaluation & Licensee Unit 2 SG Insp Results

ML20134Q136
Person / Time
Site:	South Texas
Issue date:	02/24/1997
From:	Alexion T NRC
To:	NRC
References
NUDOCS 9702260280
Download: ML20134Q136 (36)

Text

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MEMORANDUM T0: PD IV-1 File February 24, 1997 i

'FROM: Tom Alexion- ORIGINAL SIGNED BY:

SUBJECT:

LICENSEE'S UNIT 2 CYCLE.6 R00 CLUSTER CONTROL ASSEMBLY (RCCA) EVALUATION AND LICENSEE'S UNIT 2 STEAM GENERATOR (SG)

INSPECTION RESULTS  ;

-I recently received the subject faxes from the licensee. -In addition, I l had recently faxed questions from the NRC staff on SG inspection results to help structure;that telephone call with the licensee.

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l-l Docket No. 50-499 i

Attachments: 1. Unit 2 Cycle 6 RCCA Evaluation and Response to NRC Questions on Incomplete Rod Insertion

2. Uniti2:SG Inspection Results

3. Questions on Unit 2 SG Tube Inspection Results a

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DATE- k79[ '

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- WASHINGTON, D.C. 30006 4001 l

s.,*****j February 24, 1997 l-MEMORANDUM TO: PD IV-1 File FROM: om Alexion

SUBJECT:

LICENSEE'S UNIT 2 CYCLE 6 ROD CLUSTER CONTROL ASSEMBL ,

(RCCA) EVALUATION AND LICENSEE'S UNIT 2 STEAM GE INSPECTION RESULTS I recently received the subject faxes from the licensee. In addition, I had recently faxed questions from the NRC staff on SG inspection results to help structure that telephone call with the licensee.

l I

l Docket No. 50-499 Attachments: 1. Unit 2 Cycle 6 RCCA Evaluation

! and Response to NRC Questions on Incomplete Rod Insertion

2. Unit 2 SG Inspection Results

3. Questions on Unit 2 SG Tube Inspection Results I

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., .-.n- - ,, 4.. , ,,

i Plant Cg?- -Review Committee

• FORC Review Cover Sheet 41 0102- Revi ion No_

O d506 -

% n~.- No. '

mu_ 6f HL-II.83 ; (A.*+'t @A 6 M Ev4*b j

1 The PORC has reviewed this item and has L.. * --% (check as approprime):

/does E k dnes involve an UNREVIEWED SAFE 1Y QUESTION. 1

. / adversely isnpset plam nuclear safety.

It does . . - -does E does E adversely 'unpact the health and safety of plant personnel or the public.

k does  ;

it #as does E aquire fweer seview by es Piem Mr.ee NSRB. or a6er individuaWyoups.

_d __

J / NSRB Other(specify below)

Plant Manager Uni I 2 REMARKS The PORC recorarnends thisitem for:

DISAPPROVA1. UntS PORC MEENO No. f 7'O/0 -

MPfAOVA1.

COMPLETED BY DATE__ T T 1 PoRC s.6ifwy i,e n. .f ec o,isinmin, ws (nrm. . hen co c d. suAu. 6e in in.4 6. aoe.rd.nce wie e, ee.euion documem.

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ATTACHMENT 1

k. , e - -

OPGP05-ZA-0002 Rev.5 Pase 38 of 41 10CFR50.59 Evaluadoas Unreviewed Safety Question Evaluation Form (Sample) Page 1 of 4 Form 2 - -

1 Unrenewed Salety Question Evaluation e 97 0102 Rev.No. O Page l of (

Rev. No. O originatino Document sT-ue+t sess, unn 2 crese a mecA Evehntkm NOTE: Attach 10CFR50.59 Screening Form or License Compliance Review Form to this USOE.

TPNS 0.N!b N/A UNIT 1 O UNIT 2 E BOTH O System two-letter designator or structure narne NOTE: Use additional sheets as necessary to provide the bases.

A.1

1. Does the subject of this evaluation increase the probabtfity of occurrence of an accident previously evaluated in the Safety Analysis Report?

O YES S NO Bases: See Attached.

II. Does the subject of this evaluationincrease the consequences c'sn accident previously evaluated in the Safety Analysis Reprt?

O YES S NO Baset.: See Attached.

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111. Does the subject of this evaluation increase the probability of i occurrence of a malfunction of equipment important to safety previously evaluated in the Safety Analysis Report?

O YES S NO Bases: See Attached. -

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IV. Does the subject of this evaluation increase the conseques:ces of a malfunction of equipment important to safety previously '

evaluated in the Safety Analysis Report?

Bases: See Attached.

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i Page 39 of 41 OPGP05-ZA 0002 Rev. 5 f 10CFR5039 Evaluations j'

t Unreviewed Safety Question F. valuation Form (Sample) Page 2 of 4 Form 2 Unreviewed Safety Question Eva!uation # 97 0102 Rev. No. O Page 1 of q Rev.No. O

! Originating Document: st.us-Hi teas, uret 2 cyoie s RCCA av m anen

! A.2

l. Does the sub)oct of the evaluation create the poesitWty of an accident of a afferent type than any previously evaluated in the Safety Analysis Report?

O YES E NO Bases

See Attached, j

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j II. Does the sub}ect of this evaluation create the possibiltty of a different type

of maltunction than any previously evaluated in the Safety Analysis Report? S NO l O YES i Bases

See Attached.

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A.3 L Does the subject of this evaluation reduce the marDin of safety as defined l in the basis for any Technical Specification?

YES S NO l

Bases: See Attached.

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Page 40 of 41 OPGP05-ZA 0002 Rev. 5 10CFR50.59 Evaluations Unreviewed Safety Question Evaluation Form (Sample) Page 3 of 4 l Fonn 2 l Unteviewed Safety Question Evaluation # 97-0102 Rev.No. O Page 3 oi 4 ST-UtbH1.1483. Unit 2 Cycle 8 RCCA Rev.No. O Originating Dooument:

Evoluenon l

SAFETY EVALUATION

SUMMARY

'Iliis evaluation has shown that the South Texas Project has taken reasonabic precautions to ensure that the failure o the RCCAs to fully insert will not occur, or otherwise be limited, for the duration of Unit 2 Cycle 6. However, sho RCCAs becorne stuck, the Safety Analysis provides bounding resuhs with respect to the Reload Safety Analysis Checklist. Furthermore, this safety evaluation demonstrates that the subject condition of the RCCAs failing to insert given the hounding scenarios examined for South Texas Unit 2 Cycle 6 is acceptable since it does not represent an unreviewed safety question according to the criteria of 10CFR50.59.

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OPGP05-ZA-0002 Rev.5 Pase 4i of 41 100tu50.59 Fvaluadens Unreviewed Safety Question Evaluation Form (Sample) j Page 4 of 4 Form 2 Unreviewed Safety Question Evaluation # 97 0102 ___

Rev.No. O Page _ h- of Q_

Rev.No. O Originatino Document: gg.insa, una a cyole s RCcA B. 1. v All of the above questions were answered No therefore,the originatina document does nelinvolve an Unreviewed Safety Question, the Condition Report AcGon for changing A Uie UFSAR per OPGP06-ZN-0004 is _ .

2.

One or rnore of the above questions was marked YES: therefore, the originating l docurnent involves an Unreviewed Safety Question. The originating document, as presented, shall NOT be implemented without prior approval by the NRC. Provide a 4

recommendation for disposition of the Unreviewed Safety Question below. Refer to OPGP05 ZN-0004 for processing licensing amendments. Further processing of this form

! to the PORC, Plant Manager and NSRB is DQ1 required. Notify Procedure Control that

] the evaluation involved an Unreviewed Safety Question so that Procedure Control can close the USOE number.

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i l q REGOMMENDED DISPOSITION; l

Ayewve USOE CC'l- o\o 2- l

\ l PREPARED BY: 5, b (diwh [. h, VMt- '((/(( ((

~ ORIGINATOR g Date REVIEWED BY: ~

Q b. @epe--

3. L. boterQUAllFIEDREVIEWER v

~

2~ lM7 Date APPROVED BY: .

- p 2 'F/7 Data Pore -,Na No.

fPANIMENT MANAGER 17-IO Mnh7 Dale APPROVED BY:

i e[f Fe PLANT MANAGER

~/N/f7Date HEMARKS:

0 Page35er4:

OPGP05-ZA 0002 Rev.5 10CFR50.59 Evaluations 10CFR$0.59 Screcting Form (Sampic) Page I of 3 ,

Form 1 l 4 O DEStGNCHANGE @ OTHER DUNIT et O UrSAR CN St.NT #2 l

TPNS # .M% i UNIT 1 UNIT 2 E BOTH O l Bystem two-letter designator or structure name N8 REV.NO. O ST-US HL 1953. UnM 2 Cycle 6 RCCA Evaluation _

ORIGINATING DOCUMENT NO.

~DES CRPTION OF Ct4ANGE  :

Should the South Texas Unit 2 Cyeis 6 esperience the condition where, upon reactor trip, the Rod C to tuny insert,the safsty evolustion contained herein demoneirates that the outduct condicon does not inv i

l PREUMINARY SCREENING YES NO q

8 Does the proposed chante represent a change to the Plant Technical Speelecatiors? O j

! le en Unreveswed Safety Question known to be aeoncialed wtm the subject change? Q g  ;

2.

i NOTE: ll'YES' to either stuostlone 1 or 2 roter to OPOP06 ZN#04.

1 Does the proposed change represent Q g 3 A change to only correct a typogrsphical, editorial or des *,hD error?

4. A which is iconecal to end addreseed in les entirety by an existing approved 1DCFR50.59 C @

Sor SOE or NRC approved noensing submfuaf?

l

- w,; parWcomponent change v.'ah an equfvalent part/ component? O g

! 8 A spara or n '2.3 for a definitton of equivalent)

(See Secton

s. A conneuravon chanes wiwn esisting men specincomw. O E R

ff att enemers to the above guestions are *NO* perform the ftnet screeNng end rnark N/A in the approval blocke Dolow.

l et the answer to any gueston (3) through (6) le YES* a final screening is not neccesary, i

84lr. approval blocus below and t$ecard pa0es 2 and 3.

Provide a lustlRcaton and retorences if any of tems (3) through (5)le answerod YES",

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' The UFSAR requires revtslon per OPOP06 2N 00047 g

Yne Corwellon Report Action for changine the UFS AR le Npered h jhd 4

, Oste

'On.e l

l I we M/4 a" ' ' a- '

a- l l L 1

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Rev.5 Page 36 of 44 l OPGF35-ZA-0002 10CFR50J9 F.vskaations i 10CFR3039 Screening Form (Sampic) Page 2 of 3 Forrn 1 .

Rev. No. O j Originseng Documert No. ST4&HL 1643, (Jnit 2 Crete 8 RCCA Evoluedon

]

f PINAL SCREENING l the beeneing l irweeves somem that is not described in in 5AR and is notation for each In respones to N quesdons below,if tie must be documented Wth adequate technical l Is appropriate. However,lh6: attrtMos rewtowed should be indcated. The houng of I beels.

1 quesson and sectione reviewed of apppostdo documents and l 1

sepitiutes and documorits for 10CFRSO.te screering can be found in Addendum S. ,

l j " @ YES Q NO traor n yec,dien8 appropn tam' Coordines.on conounence. Required?

J Awk and Retabahy Anatye6s C ThermatHydrouhes

[Reentor Engr. @ */1/,,

O *c' O EO ~@ Omm Ametor O c8va O **ch Eno- l

_ 1 YES NO I i.

oose me ub,.ct ce ini. ree.w inva. a e.ng. = = incey = dese d in = s.iefy Annysi. E O I Repon?

UFSAR Section 4.2.3 6 states,"The guide INmbles of the fuel assernbha provide a clear channet for Inserton of the rod control rods." This could be construed me devisung from sie stme statement should the rode tell to t.Ay insert to rod bottom during unit 2. Cycle a. Therefore, the subject in review is canaldered a change to lhe facillry.

0

2. Does the subject of this review trwolve a change to the procedures as described in the Sofety Q Q AneWe Report? Refer to OPAP01 ZA4103.

Felbre of certain rods to fully insert to ved bosomInvohwe a change to the facility end does not Irwolve a ohenge to procedures.

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1 Rev. 5 Page 37 of 41 iJ OPGP05-ZA-0002 l

' 10CFR50.59 F. valuations 10CFR50.59 Semening Form (Sample) Page 3 of 3 Form 1 Rev No. O oripnesing Daoumont No. ST UBW1983 Unit 2 Cycle 6 MCCA Evatustion YES NO 1 Does the out4ect of this review propose the conduct of test or esperimente not descrbed in me g g Salsty Analyst Report?

The tature of certain rode to tuuy insert to rod bottom ie a chan0e to the faculty and does not propoor, l

j eny newlasts or esperiments.

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! 4. Does the proposed change affect Condllion6 or besos estumed in the Saf Anelvels Report or does not entall

@ Q seesty-related fanctions of eaJpment/ systems, even though the proposed eny phyolcal change in emetuig structures, systems, or procedures as descrfbe in the SAR7 J

The tellure of mutuple control rode to tutty insert to rod bottom has the poesntlet to sHect assumpsons in l

i eeveral safety Anarynes that rely on rod clutter control ensemblies for f einvity contro sor anuidown.

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\ W any anewer is afirmstrve, complete the screening form and perform en unreviewed Sefety Question Evaluation.

! N e5 answers are negasve, no Unreviewed Safety Queadon Evalueson is required.

{

i The UFSAR metAree revtelon per OPGP05 2N40047 @

The Consoon n. port Acuan for chanong me UrsAn i.

l er. pare er '3. M LJlaton 0.e. +M= M,v,:_l & 2//fr/97ae.e

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i Approved.t 17. L .%er- v m m.v.we, L Qe v 2-1 B'- 97 oate 1

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O U2C6 RCCA Evaluation Page 1 of 18

- USQE 97-0102, Rev. O Attachment to USQE 97-0102, Rev. O Unit 2 Cycle 6 RCCA Evaluation (J. M.Wigginton. February 1997)

Page Table of Contents 2

1.0

SUMMARY

2 1.11ntroduction 2

1.2 History 4

2.0 REMEDIAL ACTIONS 5

3.0 MECHANICALEVALUATION 7

4.0 SAFETY ANALYSIS 7

4.1 Impact of Incomplete Rod Insertion on Power Reduction 8

4.2 Core Physics Analysis 9

4.3 ims of Coolant Accident 10 4.5 OtherSafety Related Areas 10 5.0 TECHNICAL SPECIFICATIONS AND SYSTEM FUNCTION 11 6.0 DETERMINATION OF UNREVIEWED SAFETY QUESTION 13

7.0 REFERENCES

15 8.0 TABLES AND FIGURES 15 5.1 Table 1, Hypothetical Stuck RCCA Positions 17 8.2 Figure 1, Fuel Assembly Dashpot Regium 17 8.3 Figure 2,20 RCCAs Stuck at 16 Steps 15 8.4 Figure 3,12 RCCAs Stuck at 22 Steps

U2C6 RCCA Evaluation Page 2 of 18

- USOE 9'7-0102, Rev. O 1.0 Summary 1.1 Introduction The purpose of this evaluation is to determine the impact of this condition on the safety analysis of record and provide documentation that this condition does not seiom..; an unreviewed safety question for South Texas Unit 2 Cycle 6. 'Ihis evaluation will show that the South Texas Project has taken all EaM precautions to ensure that the failure of the RCCAs to fully insert will not occur, or otherwise be limited, for the duration of Unit 2 Cycle 6. However, should RCCAs become stuck, this safety analysis provides bounding resuhs with mapact to the Reload Safety Analysis Checklist. Furthermore, this safety evaluation demonstrates that the subject condition of the RCCAs failing to insert given the bounding scenarios examined in this USQE for South Texas Unit 2 Cycle 6 is acceptable since it does not represent an unreviewed safety question according to the criteria of 10CFR50.59.

Section 1.2 of this report presents a brief discussion of the history of the incomplete Rod l

j Insertion (IRI) issue at STP. Section 1.1 documents the occurances of rods falling to insert at STP. Following the historical review of events a discussion of the remedial actions taken to 3

' address the IRI for U2C6. Remedial actions have been taken reduce the probability of IRI during U2C6 In addition evaluations have been performed to demonstrate that if IRI occurs during i

. U2C6. adequate trip reactivity and shutdown margin exist. Section 3.0 presents a discussion of l

evaluations that have been performed conceming the IRI issue at STP. Section 4.0 provides a i discussion of the Safety Analyses performed to address the IRI issue. Discussions of the impact j of 1RI on trip reactivity and shutdown margin are presented. Section 5.0 presents a discussion of 5

the IRI issue's impact of Technical Specifications. Section 6.0 provides the responses to the 7 i USQE questions. The USQE question responses provide the basis for the determination that the .

' IRI issue does not pone an unreviewed safety question for U2C6. Section 7 provides the

references used in the development of this evaluation. Finally Section 8 contains all tables and j figures.

1,2 History i la Cycle 6 of Unit 1, it was noted that several Rod Cluster Control Assemblies (RCCAs) werc

! failing to fully insert following various plant trips and rod drop tests. Specifically, failure to

fully insert was first noted during a reactor trip on 12/18/95 when three RCCAs in core locations i

F-10, C-9, and N 7, were shown by the Digital Rod Position Indication (DRPI) system to be at 6 l steps withdrawn. The RCCAs in locations C-9 and N-7 belonged to Shutdown Bank B and the

! RCCA in F-10 belonged to Control Bank C. Stuck rods occurred in Standard fuel assemblies with accumulated assembly burnups of approximately 43,000 MWD /MTU. A prior reactor trip

on 8/29/95 was uneventful.

In subsequent rod drop testing performed on 3/2/96, the RCCAs in locations N-9, D 8, F-6 and K-10 were also observed to be at six steps withdrawn. These RCCAs were also located in Standard fuel assemblies with burnups greater than 43,000 MWD /MTU.

3 1

- - - -- ~ n U2C6 RCCA Evaluation Page 3 of 18

- USQE 91-0102, Rev. 0 When the reactor was shutdown for 1RE06 on 5/18/96, the RCCAs in locations C-9 N 9, D-8, F-6 and K-10 were again ohnerved to be stuck at six steps withdrawn while N 7 and F-10 were at 12 steps withdrawn. In addition, the RCCAs in locations C 5, C 7, E-11, and K-8 were all ahown to he at 6 steps withdrawn, making a total of 11 RCCAs which did not fully insert. All stuck rods occoned in Standard fuel assemblies with the lowest everage assembly burnup at appmximately 32,000 MWD /MTU.

Subsequent drag testing of UIC6 assemblies wu performed in July of 1996 for CR 96-14358.

During the drag testing, the assemblies that failed to fully insert on 5/1856 exhibited excessive dragla the lower dashpot region.

In rod drop testing performed during Cycic 7 of Unit 1 on 1/25/97, the RCCAs in core locations C 9 and K-8 were observed to stick at 6 steps from the bottom after being dispM. These RCCAs were located in V511 fuel assemblies with a lowest average assembly burnup of approximately 26,000 MWD /MTU. These were the first observed incomplete rod insertions in V5H fuel assemblics.

When Unit 2 was shutdown for 2REOS on 2/8/97, rod drop testing was performed. During the easting, the RCCAs in core locations D-8,E 11.F-6 and H 8 stuck at 6 steps, while the RCCA in core location F-10 stuck at 12 steps making a total of 5 RCCAs that did not insert. All stuck rods occurred in Standard fact with a lowest avesage assembly burnup of approximately 40,000 MWD /MTU.

The Unit 2 End Of Cycle 5 reOln.p lesting also indicated that several rods experienced a small arnount of slowing above the fuel usembly dashpot. 'Ihe maximum rod drop time (dashpot entry) was 1.83 seconds in a high hunup (approximately 50,000 MWD /MTU) Standard fuel assembly, which represented a 0.25 second increase since the previous rod drop test performed on 01/11/96. All drop times were well helow the Technical Specification limit of 2.8 seconds.

1he increased rod drop times are an indication of increased resistance above the fuel assembly ,

l dashpot region.

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U2C6 RCCA Evaluation Page 4 of 18 USQE97b102,Rev.O l

2.0 Remedist Actions 1

'Ihe South Texas Project has taken several actions to both minimize the possibility of this phenomena occoning in Unit 2 Cycle 6, and to ensure that adequate shutdown margin and trip ,

l mactivity emistifIRIdoes occur. )

First and foremost, the U2C6 core was redesigned to minimim the bumup experienced by assemblies in rodded locations. 'the cunent core design has only fresh V5H fuel assemblies and once burned Standard fuel assemblies from U2Cl in rodded locations. The projected maximum and of hot full power bumup for V5H assemblies is 26,300 MWD /MTU and 32,000 MWD /MTU for Standard fuel assemblies. Including a 52 EPPD coastdown, the projected maximum EOC burnup for V5H assemblies is 29,000 MWD /MTU and 34,000 MWD /MTU for Standard l

assemblies.

In addition to core redesign, all of the U2Cl assemblies that are to be loaded in U2C6 were drag seated in the U-2 Spent Puel Pool as an action for CR 97-1805. All assemblies exhibited drag l forces that were well helow the Wesunghouse F-5.1 specification criteria (less than 100 pounds in dashpot,less than 40 pounds tbove dashpot).  !

Safety Analyses have been performed which postulate RCCAs sticking in worst case scenarios, as discussed in the Safety Analysis section of this document. '!hese safety analyses bound all incomplete rod insertion conditions experienced to date for STP Units 1 and 2. Based on the facts that; a) The analyses bound previous incomplete rod insertion conditions, b) test data indicates that the probability of inecmplete rod insertions increases with increasing burnup, and c) the assembly average bumups for assemblies in rodded locations of the U2C6 core are minimized, the safety analyses will bound any incomplete rod insertion conditions that may occur for U2C6.

I U2C6 RCCA Evaluation Page 5 of 18

- USQE 97-0102, Rev. O i

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3.0 MechanicalEvaluation 1)e South Texas Pmject conducted md drop time testing for the affected banks to help establish the characteristics of the tw=i' --: md insertion for Unit 1 Cycle 6. The md dmp pmides provided important information relative to the affected RCCA locations. Specifically, the traces indicate that 4 all RCCAs inserted within normal drop time ranges of approximately 1.6 seconds which is well j within the Technical Specification limit of 2.8 seconds.1he traces also allow for evaluation of the

lower thimble tube deceleration or interfennce. Typical good locations (fully inserted) in Standard

j. fuel of Unit 1 Cycle 6, showed characteristic spring dampening occurring in the dashpot region as j the RCCA spring pack contacts the fuel assembly adapter plate (iA recoil). This spring acdon is l

clearly missing from the affecsad annemblics, indicating the failure of the RCCA to overcome the Interference within the dashpot at these locations and achieve the same free motion. All tests l indicated that the dme to dashpot enuy was not affected and is consistent with previous startup testing performed in the Spring of 1995 for Unit 1. l During the Unit 1 Cycle 6 rod drop tests in December of 1995, an additional RCCA failed to fully insent at core location N 9. This RCCA was also in a Standard fuel assembly, and stopped ,

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appmximately 6 steps above the fully inserted position. During the tests RCCAs in locations N 7, C-9, and F-10 failed once again to immediately achieve fullinsertion. However, aner approximately I hour, RCCAs N-7 and N-9 achieved full insertion without any additional assistance.

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l Purther testing was r4m,.J in which the RCCAs which failed to fully insert were rust withdrawn i 6 steps and 16 stepped in. When this was done, full 'msertion of the RCCAs was achieved. When I the RCCAs wen absequently withdrawn to 24 steps and tripped, they stopped again at 6 steps l ' withdrawn. The RCCAs residing at 6 steps (C-9 and F 10) were then manually inserted to rod bottom.

For Unit 2 Cycle 5, rod drop testing was performed on 1/11/96 during the planned electrical genermior outage. All 57 RCCAs were tested and all rods dropped to rnd bottom without any degradation in rod drop times.

When Unit I was shutdown for 1RE06 on 5/18I96. a total of 11 RCCAs did not fully insert. All stuck rods occuned in Standard fuel assemblies with the lowest average assembly bumup at approximately 32,000 MWD /MTU. .

On June 3,1996, Unit 1 Cycle 7 Mode 5 (cold conditions) Rod Drop Tests were performed in accordance with OPSP10-DM.0003 (without surveillance credit). These tests indicated nor responses (i.e. including recoil) for all rodded locations.

Drag testing was performed in July of 1996 to determine the root cause of the UIC6 incomplete rod insertions. A matrix of Standard fuel assemblies from UIC6, including both assemblies that stuck and assemblies that fully inserted on 5/18/96, went tested. During the testing, assemblies that failed to fully insert on 5/18/96 were observed to exhibit excessive drag in the lower dashpot region. The Spent Puel PoolIRI testing results indicated that the dashpot region of the

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- USQE 97-0102, Rev. O U2C6 RCCA Evaluation Page 6 of 18 assemblies that stuck on 5/18/96 had become deformed. Based on the Spent Fuel Pool IRI testing results the root cause of the UIC6 incomplete rod insertion problem was determined to be guide tube distortion in the dashpot due to the in-vessel axial compressive load. The "most likely" cause for the excessive distortion was determined to be inadequate resistance to mechanical buckling in the fuel assembly design. Contribudn5 factors included " irradiation effects and thermal creep a higher burnup levels".

'the upper guide tube clearance with the RCCA is more than 5 times that of the 6 mil diametral dashpot clearance. Any interference within the guide tubes in the high burnup assernblies is likely to be limited to the lower dashpot ases, is., in the region from 0 to 14 steps withdrawn (Figure 1). This is supported by the test resuhs performed on South Texas Units 1 and 2.

When Unit 2 was shutdown for 2REOS on 2/8/97, rod drop testing was perforned. During the testing, RCCAs in core locations D-8, E 11. F-6 and H-8 were shown by the Digital Rod Position Indication (DRPI) system to be at 6 steps withdrawn, while the RCCA in core location F 10 was shown by the DRPI system to be at 12 steps withdrawn, making a total of 5 RCCAs that did not .

fully insert. All 5 RCCAs were inserted into Standard fuel assemblies, with a lowest average i assembly burnup of approximately 40,000 MWD /MTU. j 1he Unit 2 End Of Cycle 5 rod drop testing also indicated that several rods experienced a small amount of slowing down above the fuel assembly dashpot. The maximum rod drop time (dashpot entry) was 1.83 seconds in a high bumup (9 50,000 MWD /MTU) Standard fuel '

assembly, which represented a 0.25 second increase since the previous rod drop test performed on 01/11/96. All drop times were well helow the Technical Specification limit of 2.8 seconds.

1he increased rod drop times are the an indication of increased msistance above the fuel assembly -

dashpot region. The increase in rod drop time is not expected to impact U2C6 since a rod drop time increase of this magnitude has only been observed in a high burnup fuel assembly U2C6 was adesigned to minirnize assembly bumup in rodded locations.  ;

1he above evaluatiom support sevent important conclusions:

e lhe interference or binding causing incomplete insertion is limited to the lower dash pot agion.

• Scram times are well within Technical Specification requirements, e lhe incomplete rod insertions are a reant occurrena as the prior scrams (before 12/15/95) were uneventful.

• The test results did not indicane that the phenomenon was related to a loose part.

• 1he test results do not indicate that the phenomenon is related to degmdation in the RCCAs.

-- - e U2C6 RCCA Evaluation Page 7 of 18 USQE 97-0102, Rev. O 4.0 Safety Analysis Westinghouse has performed Core Physics Analyses for the various conditions listed in Table 1 and Figures 2 and 3 for Unit 2 Cycle 6.1hese analyses show that all Reinnd Safety Analysis Checklist (RSAC) g.ger. continue to be met for a variety of RCCAs stuck at various core locations and positions (as depicted in the Table and Figures). Key among these F- ;ers is Shutdown Margin and Trip Reactivity,which continue to be satisfied. This demonstrates that 10CFR50 Appendix A, General Design Criseria will ocatinue to be satisfied, c:recially the following:

• ODC 25 Protection system requirements for reactivity control malfunctions.

• GDC 26 Reactivity control system redundancy and capability.

• GDC 27 Combined reactivity contml systems capability The contml rods pmvide two functions with regard to the safety analysis. First, the control rods must insert sufficient reactivity to reduce power such that safety limits (DNB and reactor pressure) are not exceeded (trip reactivity). Secondly, the contml rods must insert sufficient negative reactivity to maintain shutdown margin as defined in Technical Specifkations 3.1.1.1 and 3.1.1.2.

4.1 Irnpact ofincomplete Rod Insertion on Power Reduction The trip reactivity requirement for the full power accidents def'med in UFS AR Charter 15 is satisfied by all but the highest worth RCCA falling into the dash pots within the Technical Specification rod drop limit of 2.8 seconds dudng an accident. During the accidents the highest worth rod is modeled to remain full out. Since all measured rod drop times (dashpot entry) to l date are well below the 2.8 second limit, and since no rods have been observed to stick above the dashpots, the trip reactivity requirement for full power accidents will continue to be met for U2C6.

I i The Trip reactivity requirement for the Rod Withdrawal From Subcritical (RWFS) accident for  !

i U2C6 is 2 % Ap. Westinghouse performed bounding trip reactivity evaluations for U2C6 based on all inserted rods stuck at 28 steps for without a post-trip cooldown or stuck at 22 with a post l j

trip cooldown from 567'F to 550*F. The results of the evaluation are presented in Table 1.

]

Table I shows that the RWFS trip reactivity requirement is met in both cases.

In order to evaluate the IRI safety implications related to shutdown margin, shutdown margin 1

! evaluations were performed based on the following conservative assumptions:

\

I

• All inserted RCCA: could be stuck at 12 steps or less withdrawn, or up to 12 RCCAs could be stuck at 22 steps or less withdrawn; or 20 RCCAs could be stuck at 16 steps i .

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U2C6 RCCA Evaluation Page B of 18

- USQE 97-0102, Rev. O

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withdrawn following reactor trip with or without a post 4 rip cooldown from 567'F to 550*F.

• No location or burnup restrictions were assumed in determining the number of allowable incompletely inserted RCCAs.

e 'Ihe highest worth RCCA is stuck fully withdrawn.

These assumption are conservative for the following reasons:

• Use of 12 or more (See Table 1) RCCAs represents a bd5 case since a maximum of 11 RCCAs in Unit I and 5 in Unit 2 have been shown to be affected based on numerous rod drop tests.

e The number of allowable incompletely inserted RCCAs was determined without placing any restrictions baned on location or assembly average bumup. The locations used in the analysis (see Figures 1 through 4) were selected to minimize the N 1 rod wonh (thus minimizing the available shutdown margin).

• The use of an assumed insertion to only 22 steps withdrawn exceeds the elevation at which RCCAs became stuck in actual rod drop testing.

• During Unit 2 Cycle 5, there was over 30% margin to the rod drop time limit of 2.8 seconds stipulated in the Technical Specifications.

• STP U2C6 core loading pattem has been developed to minimize rodded, EOL, fuel burnups, for example 21 once bumed Standard fuel assemblies from U2Cl and 36 fresh V5H assemblies are under control rods.

The results of the analyses for the above scenarios are presented in Table 1. Table I shows that ,

the shutdown margin RSAC limit of 1.3% Ap is met. In addition, Westinghouse determined that the case of all 21 RCCAs located in the once burned Standard fuel assemblies sticking at 14 or fewer steps with all other rods fully inserting would result in no RSAC violations. This is l

significant since there have been no recorded cases of RCCAs sticking in fresh assemblies at STP, and all stuck rods have stuck at an indicated position of 12 steps or less.

l 4.2 Core Ptsysics Analysis Specific analysis of the conditions described in Section 4.1 were performed to cover a range of cycle burnup from 0 MWD /MTU to 22,100 MWD /MTU (The end of full power operation plus coastdown). 'Ihe calculations confirmed that ALL Reload Safety Analysis Checklist (RSAC)

(sefer to USQE 97-0101) cunent limits continue to be met including shutdown margin and the current licensing basis RWFS trip reactivity limit of 2 %Ap.

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4 USQE 9'l-0102, Rev. O U2C6 RCCA Evaluation Page 9 of 18 l .

A3 Loss of Coolant Accident 43.1 LargeBreakLOCA ne Imge Break IDCA analysis does not take credit for any RCCA insertion to shutdown the com.

De void formation during the thCA produces the negative reactivity to shutdown the core.

Derefore, this accident scenario bounds the assurned operating condition.

43.2 SmallBreakIACA The Small Break LOCA analysis credits the shutdown of the core due to the reactor trip and the rod insertion time. Given that the RCCAs will continue to shut down the core within the RSAC limits and do not cause the time to go above the safety limit for rod insertion,the Small Break LOCA analysis is unaffected by this assumption.

433 ImgTermCooling The South Texas Project Long term cooling analysis does not take csedit for RCCA insertion to keep the core subcritical. De baron concentration of the ECCS produces the negative reactivity to keep the core subcritical. %erefore, this accident scenario bounds the assumed operating condition.

43.4 Hot Leg Switchover Analysis De hot leg switchover analysis does not credit any rod insertion to keep the core subcritical. De boron concentration of the FICS and the rate of safety injection during hot les recirculation produces the negative reactivity to keep the core sutx:ritical.nerefore, this accident scenario bounds the assumed operating condition.

44 Energency0penttingProcedurne ne EOP procedure actions in OPOP05-BO-ES01 requires emerpacy boration of 228 ppm for each stuck rod greater than 18 steps when more than one rod is stuck, and 60 ppm for each rod stuck less than or equal to 18 steps. De failure of the RCCAs considered in this USQE are bounded by the 228/60 ppm /RCCA criteria, since they were based on previously bounded cases as determined through the 10CFR50.59 process. Derefore, the EOPs are not iiW by this USQE.

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l USQE 97-0102, Rev. O U2C6 RCCA Evaluation Page 10 of 18 4.6 OtherSahrtyRelated Areas Other safety related areas have boca reviewed and it was determined that none of these am affected by the subject condition. These include the following (as discussed in Reference 7.1):

- mA=bal and fluid systems

- instrumentation and control systems

- LOCA and steam line break mass / energy release and impact on containment analysis

- radiologicalcanaequences

- steam generatortube rupture

- pmbebilistic risk assessment

- Technical Specirsations

- pmtection system setpoints I 5.0 Technical Specifications and System Function t Technical Specification 3.1.3.1 requires that all full-length shutdown and control rods shall he OPERABLE and positioned within +/- 12 steps (indicated position of their gmup step counter as cee.,,-cd to demand position). A control rod is considered OPERABLE if it is capable of being l tripped, movable, and properly aligned. De RCCAs have been demonstrated to be capable of being tripped, movable, and proper alignment can be maintained. De RCCAs are fully capable of performing their design function in conformance with the definition of OPERABIROPERABILITY. De accident analyscs assume that the RCCAs insert adequate negative reactivity within the prescribed drop time (2.8 sec.). Dis analysis shows that Shutdown Margin will continue to be met for those cases presented in Tr. ole 1 and Figures 2 and 3. The rod 7

drop times are still well within the required limit. Consequently, the RCCAs are fully OPERABLE j with regard to TS 3.1.3.1 and 3.1.1.1 (Modes 1,2, and 3 at no load Tavg).

! STP has concluded fmm the evaluation above that the IRI condition does not affect the Technica!

Specifration requirements for the operability of the contml rods.

i De condition will not affect the rod control system function during normal operation of the plant.

i

De RCCAs have been demonstrated to be fully contmllable by the rod control system. De

condition will have no impact on the manual or automatic features of the rod control system.

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i USQE 97-0102, Rev. O U2C6 RCCA Evaluation Page 11 of 18 j . ,

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4 6.0 Determinetlon of Unreviewed Safety Question he subject condition of the failure of RCCAs to fully insert to the rod bottom position has been f

j evaluated using the guidance of OPGP05-ZA-0002 and, on the basis of the following justification,

does not involve an unreviewed safety question per the criteria of 10 CFR 50.59.

4 i

j L gjjl _he t o# *lHev of an =L" w,*:==1v ev=1a==A in the UFSAR be irc----4?

4 Dis safety evaluation documents that the probability of an accident previously evaluated in the j UPSAR is not increased. All applicable design criteria and all pertinent liamsing basis acceptance criteria are met. De demonstrated adherence to applicable standards and i

j evance critaria precludes new challenges to components and systems that could increase

the probability of any previously evaluated accident. He fuel clad integrity is maintained and j the structural integrity of the fbel rods, fuel assemblies, and core is not affected. The condition i does not impact fuel rod performance or dimensional stability nor will it cause the core to i operate in excess of pertinent design basis operating limits. He condition does not introduce

! any accident initiators. Control rod accidents involve rod ejection or uncontrolled withdrawal, j neither of which are affected by this condition. Therefore, the probability of occurrence of an

accident previously evaluated in the UFSAR has not ir .a.c.i i

4 4

2. Will the == ==== of an acewas rsviousiv evalawai in the UFSAR be irc=--4?

j This safety evaluation documents that the consequences of an accident previously evaluated in l the UFSAR is not increased. All applicable design criteria and all pertinent licensing basis l

rv-~ criteria are met. The demonstrated adheterr.e to these standards and criteria

j. precludes new challenges to s,ir,,c.r.;;as and systems % could: a) advermly affect the ability

of existing components and systems to mitigate thr. consequences of any accident and/or, b)

J advarsely affect the integrity of the fuel rod claddinJ as a fission pmduct barrier. Furthermore. '

adhercacc to applicable standards and criteria ensures that these fission product barriers l

maintain design margin to safety. He condition has no impact on chemical, physical or mechanical properties nor will it cause the core to operate in excess of pertinent design basis operating limits. Thus, fuel clad integrity is maintained. He RCCAs are capable of

performing in ar. sod r.cc with the accident analysis assumptions for drop time and insertion of

{

negative reactivity. Since the conclusions of the UFSAR remain valid, the consequences of 1 accidents previously evaluated in the UPSAR have not increased.

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USQE 97-0102, Rev. O U2C6 RCCA Evaluation Page 12 of 18 l

j 3. May the - *ility of an arr*Aa' which is different fmm any alreadv in the UFSAR be

=

i Gualsdl 4

This safety evaluation documents that the possibility of an accident which is different imm any already in the UFSAR is not created. All applicable design criteria and all pertinent licensing l

basis Esy == criteria ese met.1he demonstrated adherence to these standards and criteria precludes new challenges to -irpra.ts and systerns that could intmduce a new type of accident. All design and performance criteria will continue to be met, no new single failure mechanisms have been created, and the core will not operate in excess of Sct design basis operating limits. The condition does not intmduce any accident initiators. Control rod AM involve rod ejecdon or uncontrolled withdrawal, neither of which are aNected by this condition. Therefore, the possibility of an accident of a different type than any previously evaluated in the UFSAR has not been created.

4. Will the o#+bilitv of a malfwtion of aaninrnant Imnortant to ..r.sv oreviousiv eval"='ad in the UFSAR be increaeM?

This safety evaluation documents that the probability of a malfunction of equipment important to anfety previously evaluated in the UFSAR is not increased. All applicable design critoria and all pertinent licensing basis acceptance criteria are met. Demonstrated adharence to applicable standards and aw.sce criteria precludes new challenges to components and systems that could increase the pmbability of any previously evalualed malfunction of equipment important to safety. No new performance requirements are being imposed on any system or component such that any design criteria will be exceeded nor will the condition  ;

cause the core to operate in excess of pertinent design basis openting limits. No new modes or limiting single failures have been created by the condition noted above. Therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR has not increased.

5. Will the gr7= == of a malfunction of mouinmane imnortant to safetv omvionniv j

evalaarad in the SAR be in===ad?

1his safety evaluation documents that the consequences of a malfunction of equipmem important to safety previously evaluated in the UPEAR are not ired All applicable design criteria and all portinent licanning basis =~=p*=na crikaria are met. The demonstrated adherence to these standards and criteria precludes new challenges to components and systems that could: a) adversely affect the ability of naisting w.uye.w. and systems to mitigate the consequences of any accident and/or; b) adversely affect the integrity of the fuel rod cladding as a fission product barrier. Furtl .veir., adherence to applicable standards and criteria ensures that these fission product barriers maintain design margin of s.fri.1he condition does not change the performance requimments on any system or component such that any design criteria will be essended nor will it cause the core to operate in excess of pertinent design basis operating limits. No new modes or limiting single failures have been created by

. - ..~ . - - - -- .

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USOE 97-0102, Rev. O U2C6 RCCA Evaluation Page 13 of 18 f . .

1 4

the condition mentioned above. Therefore, the consequences of a malfunction of equipment l l important to safety previously evaluated in the UFSAR have not ir.a r.rd.

6. May the e==thility of a ==1fa=riian of andoment imnariaat to maraev different imm any 4

ahe.afv eval ==*-A in the UFSAR be cr=*A?

2 e

This safosy evaluation documents that the possibility of a malfbnction of equipment i.sguet j to safety different fiorn any already evalussed in the UPSAR is not created. All applicable j design criteria and all pertinent licensing hasis =~*pt- criteria are rnet, ne demonstrated i

adherence to these standards and crheria precludes new chauenges to sgr.F,r r,ts and systems that could introduce a new type of a malfunction of equipment important to safety. All original l

performance criteria continue to be met, and no new failure modes have been created for any l l

! system, component, or piece of equipment. No new single failure mechanisms have been 3 introduced nor will they cause the cose to operate in excess of pertinent design basis operating i limits %erefore, the possibility of a malfunction of equipment important to safety of a l

j different type than any previously evaluated in the UFSAR has not been created.

L i 7. Will the ==ai of ""v as defiaad in the BWR to any eachateat maariftentions be i j teduced?

! I This safety evaluation documents that the margin of safety as defined in the Bases to any

} Technical Specifications is not reduced. All applicable design criteria and all pertinent i licensing basis acceptance criteria are met, especially with aspect to Shutdown Margin and j Trip Reactivity. Shutdown Margin is controlled by Technical Specifications, and this change does not reduce the Shutdown Margin below the acceptable Technical Specification I.imit.

j l k has been determined that the design and safety analysis limits remain applicable, and that l

j these limits are supported by the applicable Technical Specifications. De evaluation of this l i

condition takes into consideration normal core operating conditions allowed in the Technical l

i Specifications. His condition has been evaluated using approved design methods. His evaluation includes consideration of the core physics analysis peaking factors and core average l j

linear heat rate effects. Therefore, the margin of safety as defined in the Bases to the Technical

! Specifications has not been reduced.

i-i 7.0 References l

} References Usedin Evaluation i

j 1) ST-UB-HL-1538, " Westinghouse Sefety Evaluation Checklist (SECL 96 089), Pailure of RCCAs to Achieve Pull lasettion After Reactor Trip?, Revision 0 i

2) ST-UB-H1 1539, " Westinghouse Safety Evaluation Checklist (SBCL 95 192), Pallure l of RCCAs to Achieve Fullinsertion AAerReactor' Dip *, Revision 1

)-

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USQE 97-0102, Rev. 0 ,

U2C6 RCCA Evaluation Page 14 of 18

3) ST-UB-HL 1536," Westinghouse Safety Evaluation Checklist (SECL 95-192), Failure )

of RCCAs to Achieve Full insertion Aher Reactor Trip.", Revision 0 l

4) CR 95-14358," Unit I Control Rod Lucrtion Anomaly."

5) ST-UB HL-1683," Westinghouse Safety Evaluation Checklist (SECL 95-192), Failuse l of RCCAs to Achieve Full Insertion Aikr Reactor Trip.", Revision 4 References Checked .

6) NUREG 0781, Safety Evaluation Report for the South Texas Ptoject (Sections 4.2,4.3, t and 15) .

I

7) Updated Final Safety Evaluation Report (Sections 3.1,3.9,4.2,4.3,7.7, and 15)

8) USQE 97-0101. " Unit 2 Cycle 6 Reload Safety Evaluation," Rev. 0

9) 5N079NB1000 " Accident Analysis Design Basis Documents" l 10) Title 10, Code of Federal Regulations, Part 50 Appendix A, General Design Criteria l

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USQE 97-0102, Rev. 0 U2C6 RCCA Evaluation Page 15 of 18 8.0 Tables and Figures 3.1 Table 1, HypotheticalStuckRCCA Positions VL- . . . .

L. .. AA $L.M ': 1 + 1 h.

South Texas Project Unit 2 Cycle 6 Without Post Tr p Cooldown Stuck Trip Cycle Bumup Stuck Rod Core Case Position No.of Shutdown Reactivity * (MWD /MTU) Location Margin **

(Step Rods Withdrawn) l I l 1 22 12 l 3.17 % 2.19 % l 150 l H8 2 16 20 1 3.30 % l 150 l H8 3 12 56 l 3.24 % l 150 j HR 4 22 12 l 2.18 % 2.67 % l 22100 l H8 5 16 20 l 2.18 % l 22100 l H8 6 12 56 l 2.03 % l 22100 i HH

.. 2E i South Texas Project Unit 2 Cycle 6 With Post Trip Cooldown(550*F) 1 22 12 3.03 % 2.38 % 150 H8 2 16 20 3.16 % 150 HB 3 12 56 3.09 % 150 H8 4 22 12 1.66 % 2.86 % 22100 H8 5 16 20 1.65 % 22100 H8 6 12 56 1.50 % 22100 H8

• The trip reactivity calculations were performed assuming all inserted rods at 28 steps without a post-trip cooldown, or 22 steps with a post trip cooldown.

• • The cycle bumup of 150 MWD /MTU corresponds to BOC conditions. The cycle bumup of 22,100 MWD /MTU corresponds to EOC conditions (i.e. the end of hot full power (555 EFPD) plus a 52 EFPD cuantdown).

The RCCAs incompletely inserted in this analysis were selected to minimia the inserted rod worth. The listed shutdown margin values are therefor: representative for any group containing the indicated number of RCCAs stuck at the indicated number or fewer steps withdrawn, including the highest stuck rod fully withdrawn. The shutdown margin and trip reactivity RSAC limits are 13% and 2.0% Ap, respectively.

In addition, the analysis results indicate that the case of all 21 RCCA: in non-feed (Region 1) assemblies in the Cycle 6 core become stuck at 14 steps or fewer withdrawn would result in no RSAC violations.

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USQE 97-0102, Rev. 0 -U2C6 RCCA Evaluation Page16of/$ l 18 0.2 Figure 1, Fuel Assembly Onshpot Region 1

i LOWER GUIDE TUBE GEOMETRY

,..., s. s - -

leo m Top Nozzle

,_ 33.5 in O.450 ID Start top Jaohpoc i 38 ot*P* I O.397ID l

22in i 21.2 M End top dashpot "N '

GM #2 ll NH(ll !l$$$$$MI N

em St. % 1 14088P' i i I O.45010 lahw

• d ecepe eiri(~4habam

., zod not:4m)d esere n.2 m

+5otspe  ;

l 936lrs j

ltad Sottom - O etape 0.58710 l 3.h h hk bOll ink N hN L5 h g

Start of setae fuel op of ($ottom Nozzic,,,

!' g

_ . . - = . . . - - - . . - . - . - . . - . - - .

,i .

l USQE 97-0102, Rev. O U2C6 RCCA Evaluation Page 17 of 18 8.3 Rgure 2,20 RCCAs Stuck at 16 Steps R P N M L K J H G ,

9 E D C S A

!  ! l l l l i a l SA l lB SB lC lBl l SA l j s } SD SB SC 4 SA SP SP SA L' s SC SD

~ ~ -

e T SP T 7 SB SB SP e C D C e SP SB SB so B SP "

SP B

__ l si SD SC u SA SP SA is SC SB SB SD C B SA 14 l SA l lB l l

, Rank IdentIfler# of Locations A 8 B 8 C 8 D 5 SA 8 i SB 8 l Sc 4 l SD 4 i SE 4 SP (Spare) 12 Shaded locations indicate stuck RCCAs at 16 steps; RCCA H8 is assumed to be stuck fully withdrawn.

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USQE 97-0102, Rev. O U2C6 RCCA Evaluation Page 18 of 18 l

B.4 Mgure 3,12 RCCAs Stuck at 22 Steps j R P N M L K J H G F E D C B A s SA lB lCl lBl SA SD l SB l l SB l l SC SA D D SA 4

l SP l SP l SC A A SD s

B SP SP B e

SB SB SP y

C D C a

SB SB e SP B SP SP B to is SD A A SC SA D D SA se l SP l SP l is SC l SB l l SB l l SD

$4 SA B C

]Bl SA Rank Tdantifier s of Locations .

A 8 B B C 8 D 5 SA 8 SB B SC 4 SD 4 SE 4 SP (Spare) 12 Shaded locations indicate stuck RCCAs at 22 steps; RCCA HB is assumed to be stuck fully withdrawn.

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STP Response to NRC Questions on Ta~=alete Rod Insertion Anomaly at 2RE05 (

i i -

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1. Are there any plans to i.e.Jonn tod drop testing prior to operation of Unit 2 Cycle 67 i

Yes, rod drop time testing is nquiredprior to reactor startupfollowing core alteration

) perfonned when RCS Tavg is greater than 561*F withfour reactor coolantpumps

• in addition, weplan to collect rod drop data at cold noflow condstions when ne control rods are tripped into de corefrom the rqid rgfuelingposstion.

i 2. How are you going to i--

~+ that the control tods will provide adequate shutdown margin tad j

trip reactivity through and of core life? How do you assure that the control rods are tripp

(

. Followkg the Unit 1 January 25,1997 roddrop testing, a core loading pattern revisi i

Unit 2 Cycle 6 was initiated. The newpattern limits rodded core locations tofresh V5 H oncesurned non-V5HJkel with end-of-cycle burnys at approximately 26.6 and 32.1 l ,

gulhntu, respectively. A roddrop test severalmonths before end-of-cycle (correspondi i the lowestfuel burnsps observed in V5H and non V5H)kel) may be necessary to ensure sqfety limits are met. We will evaluate de results of Unit 1 Cycle 7 rod drop testing to l conprm that adequate sqfety margin exists, and detennine the schedulefor additional rod drop testing ($any) by mid-November 1997.

Core design calculations willbe wrformedprior to starty to demonstrate that a bounding number ofcontrols rods, independent offuel assembly burnup, which stick at 10 steps,16 steps, or 22 steps, asssasing As most reactive stuck rodfully withdrawn, will meet shutdown margin and trip reactivity safety analysis requirements through end ofcors life. STP will make Aese cole =1^"ns anulable to Ae NRC. 2hese results willbe usedin a docum  ;

safety evaluation, and will be completedprior to r'sactor starty of Unit 2, which is curren{

scheduled to occur on February 24,1997. Dnis sqfety evaluation willalso beprovided to the NRC upon STP managementapproval. i Rod drop testing performed during Unit 2 Cycle 5 demonstrated that rod drop times and shutdown margin will remain within Technical Specsycation limits, and that the rods arefree to insert into the dashpot region win the IRIcradition. During reactor operatson, a monthly rod exercise test is performed to assure that the rods are movable, and thus assumed to be

\

tttppable.

3. What will you do if rod drop tests prior to startup indicate no recoilh Recoildatafrom Unit 1 Cycle 6 and Unit 2 Cycle 5 initialstarty testing demonstrated that low recoil (0 or 1) at BOC did not impede the qWected control rod locations to perform their sqfetyfunction Arough EOC. This will be one of thefactors used to consideradditional rod drop tening describedin Question #2's response.

teeze*d 0628 E46 EIS mis e 17 W3 m PE:ET M61-*BM

, ur nesponse to NR i

4. Provide an cverview of fuel ersminations during the 2RE05 refueling outage.

Due to a high drag condition observedduring Spent FuelPool shuffing ofo hafnium control rodsfrom one of the once-burned reload assemblies, con testing ofall rodded once-burned reload assemblies will be completedprior to core r A drag criteria of100 lbs ks the dashpot and 40 lbs above the dashpot will be used to evaluate rodinsertion capability.

During the core oploaf each reloadfuelassembly willbe visually inspected with In addition, relativefuel assembly axial growth will be measured during the ofl indicated by the refueling machine n-tape when the assembly isfull down in th upender.

5. Compare appropriate parameters of std, XLR, and V-SH for 14ft core. These thimble tube and dashpot outside diameter; (b) wall thickness; (c) clearance; (d) ty ,

I XLstd YTR XL V5H Thimble Tube OD(in) 0.484 0.484 0.476 Dashpot OD(in) 0.421 0.421 0.421 WallThickness(in) 0.017 0.01 7 0.017 Clearance (in) ,

above dashpot \

0.035 0.035 0.031 dashpot 0.006 0.006 0.006 Materials l Midgrids Incone! Inconel Drc Top / Bottom grids Inconel Inconel inconel Thimble tubes Zire Zirc Zirc '

Comprehensive information is also provided in NRC Accession number % 02 010230,

" Meeting Summary"fued by Tom Alexion in the NRC Public Document Room, January 1996.

6. Provide the overall fuel length for std, XLR and V-5H fuel.

XL std XIR XL V5H Length (in)

Toppad/Botpad 188.795 188.793 188.793 gg,gg.a este Ets Eis oNistaan w370nN t'E:ET 4661-U-833

j o STEAM GENERATOR BACKGROUND DATA y SOUTE TEXAS PROJECT Unit 2 Background Information

5 2 EFPY j  ! Last cycle 439 EFPD Model E SGs - Alloy 600 MA tubing - Hydraulically expanded a.

Stainless TSP with drilled holes Thot 620 degrees Shot peaned after first cycle (NL) and second cycle (CL)

U Bend Heat Treatment R1 and 2 prior to operation Previous plugging less than .5%

2RE04 (100% TTS RPC and 22% Bobbin Coil 1 11 tubes were plugged (No circ cracks and 8 were DSIs

confirmed to be SAIs)

Unit 1 2ackground Information and Differences From Unit 2 r 4.9 EFPY at last inspection Fall, 1996 Last cycle 387 EFPD

, Wechanical Hard Rolled Tubing  ;

Carbon Steel drilled hole TSP Roto Peened prior to operation j

U Bend Heat Treatment R1 and 2 at first. refueling 1 Volt APC licensed i

l Total plugging less than 1.5%

i j 1RE06 (100% TTS RPC, 100% Bobbin Coil, Row 1 and 3 + Point) 95 tubes were plugged (56 were OD TTS Cire indications, 10 TTS j axial indications, 885 DSIs with 6 2 1 Volt) 5

)

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ATTACHMENT 2

9040 'd 8628 246 ETS Wigm]7 e n EE

ET 466T-6T-EM

i

! SCOPE OF EXAMINATIONS 2RE05 i

i I i

• All four steam generators were examined. I i

• j Bobbin coil examination of all in service tubes over their full length using a EB 810/690-LLMC (Westinghouse long life) probe.

i

• MRPC of all bobbin coil detected DSI (distorted support Indications),  !

NQI (non<quantifiable indications), all other I. codes, and all inner '

t diameter reductions at tube support plates with signal greater than

{ 5.0 Volts using a 0.115MRPC410/580 3C42PH (Zetec 3. coil MRPC)

{ probe.

MRPC examination of all in4ervice tubes at the hot leg top of i

tubesheet using a 0.115MRPC410-3C42PH (Zetec 3-coil MRPC) probe.

Approximately 100 selected tube locations (ie.; preheater baffle plate expansions, previously detected indications, etc.) were examined by MRPC using various probe designs suitable for the location geometry.

9040*d 8628 246 ETS ONISN3017 W37DnN EE:ET 466T-6T-83d

• O o e 4,

y -

3 a

R Tubes Plugged During 2RE05 Outage m

M m

Steam DSI/(sal or MAI) DSI/(SAI or mal) : Other Axial

" Circumferential Volumetric Preventative Total g Generator < = 1 Volt > 1 Volt Indications Indications Indications r

I r

A 113 7 2 0 0 3 125

B 154 4 0 0 10 1 169 C 143 1 0 0 2 0 146  ;

D 159 0 0 0 2 0 161  ;

E 601 m

b v

l

] '

Ld I i A  :

5 s .

8 7

m Y '

Si  :

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• Scuth Texts PrcDet Unit 2 2RE05 Eddy Current Daily Status Report FINAL A B C D TOTAL COLD LEG BOBBIN PROGRAM

1. PLANNED TESTS 4843 4847 4840 4845 19375

2. ACQUIRED TESTS 4843 4847 4840 4845 19375

3. RESOLVED TESTS 4843 4847 4840 4845 19375 HOT LEG BOBBIN PROGRAM

1. PLANNED TESTS 240 240 240 239 959

2. ACQUIRED TESTS 240 240 240 239 959 L3. RESOLVED TESTS 240 240 240 239 959 l

HOT LEG TUBESHEET RPC PROGRAM  !

1. PLANNED TESTS 4843 4847 4840 4845 19375

_2. ACQUIRED TESTS 4843 4847 4840 4845 19375

3. RESOLVED TESTS 4843 4847 4840 4845 19375 HOT LEG SPECIAL INTEREST RPC PROGRAM '

1. PLANNED TESTS 220 377 278 278 1153 g __2. ACQUIRED TESTS 220 377 278 278 1153

3. RESOLVED TESTS 220 377 278 278 1153 COLD LEG SPECIAL INTEREST RPC PROGRAM

1. PLANNED TESTS 46 60 49 35 190

2. ACQUIRED TESTS 46 60 49 35- 190

3. RESOLVED TESTS 46 60 49 35 190 Distribution: I Chet McIntyre, Houston Lighting And Power

• Kevin Miller /Ed Belizar, Westinghouse Copy to Data Management files Status as of February 19,1997 at 0200 PAGE 1 OF 16 90/SO*d 8628 246 EIS ONISN331l 803'13ON EE:EI 466T-6I-83d

. __ _ . - . _ . _ _ . . . --- ---- --- ~~ ~~ ~

90 d M 01 l 1- -

9 l

South Texts Proj

ct Unit 2 1 2RE05 Eddy Current Daily Status Report t -

FINAL 4

A B C D TOTAL BOBBIN PROGRAM - BOTH LEGS

1. TUBES W/INDS 1-19% TW 7 6 6

2. TUBES W/INDS 20-39% TW 9 28 ~ !

0 1 0 1 2 i.

3. TUBES W/INDS > => 40% TW 0 0 0 0 0

4. TUBES W/DSI < := 1.0 VOLTS 165 283 219 227

894

5. TUBES W/DSI 1.0-2.85 VOLTS 7 4 1 0 12 i

6. TUBES W/DSI > 2.85 VOLTS O O O O O

} 7. TUBES W/OTHER l CODE INDS i

12 36 10 6 64 HOT LEG TUBE 8HEET RPC i- 1. TUBES W/AXfAL INDS 0 0 0 0 0

2. TUBES WITH CIRC & MMI INDS O O O O O j

i

3. TUBES WITH VOLUMETRIC INDS 0 2 0 0 2 e

i _ HOT LEG SPECIAL INTEREST RPC i 1. TUBES W/ AXIAL INDS 121 158 144 159 582

_2. TUBES WITH CIRC & MMI INDS O O O O O

j. 3. TUBES WITH VOLUMETRIC INDS O 4 0 0 4 j COLD LEG SPECIAL INTEREST RPC

1. TUBES W/AXlAL INDS 2 0 0 0 2

2. TUBES WITH CIRC & MMI INDS O O O O O

3. TUBES WITH VOLUMETRIC INDS 1 4 2 2 9 o

I l

l Status as of February 19,1997 at 0200 PAGE 2 OF 16 90/90'd 8628 246 2iS DNISN3DI7 W31Tt4 PE:EI L66I-6I-83d

9 Steam Generator Tube Inspection Results l f.icensees' steam generator (SG) tube eddy current (EC) inspections play a vital role in the management of SG tube degradation. The results are used to demonstrate adequate structural and leakage integrity of the SG tubes for both condition monitoring (i.e., the as-found condition of the tubes demonstrate adequate integrity was maintained during the previous cycle) and operational  ;

assessment (i.e., the projected condition of the tubes is such that adequate  ;

integrity will be maintained during the upcoming operational cycle). 1 i

Specific information that facilitates staff reviews of licensees' condition  !

monitoring and operational assessments includes:

Primary to secondary leakage prior to shutdown Results of secondary side hydro For each steam generator, provide a general description of areas examined; include expansion criteria and specify type of probe used in each area l

For analyzed EC results, describe bobbin indications (those not examined with RPC) and RPC/Plus Point /Cecco indications. Include the following information:

location, number, degradation mode, disposition, and voltages / depths / lengths of most significant indications.

Describe repair / plugging plans j Discuss previous history; "look backs" performed Discuss new inspection findings Describe in-situ pressure test plans and results, if available; include tube selection criteria Describe tube pull plans and preliminary results, if available; include tube selection criteria Assessment of tube integrity for previous operating cycle (condition monitoring)

Assessment of tube integrity for next operating cycle (operational assessment) l Provide schedule for steam generator-related activities during remainder of current ouh;;e Hote: Licensees should be prepared to respond to the above information l during the teleconference. The staff prefers to receive responses i (e.g., simple tables and figures) to the above information prior to the teleconference.

ATTACHMENT 3

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