ML20199C056

Summarizes 971023,28 & 1104 Telcons W/Nrc & Licensee Re Independent Corrective Action Verification Program

ML20199C056
Person / Time
Site:	Millstone
Issue date:	11/10/1997
From:	Curry D AFFILIATION NOT ASSIGNED
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUM2-PPNR-0768, NUM2-PPNR-768, NUDOCS 9711190201
Download: ML20199C056 (16)

Text

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' PARSONS Daniel L. Curry,vae twwit tuw sevm Peses Dugy & Crmucats Gmsp loc.

ms unganioun naad. neaoingfennsevaw um. (cioi sssa o. F ai (cio> essa November 10,1997 Docket No. 50 336 Parsons NUht2 PPNR 0768 L U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 205$$

Millstone Nuclear Power Station Unit No. 2 IndependenLCeJ1csthe ActionYerifwatield'.tegantf1CAVP)

Gentlemen:

"Ihis lei :r transmits summaries of telephone conferences i Tween Parsons Power Group Inc., the U. S.

Nuclear Regulato:y Commission, NNECo and NEAC o.

atober 23, October 28, October 30 and November 4,1997. Tle purpose of these telephone conferences were as follows: October 23 -

Containment Spray Cale., Appli. of Safety Guide 26/ Reg. Guide 1.26, Use & Control of SF Forms, Spec.

Revision, Jumper Device index No. 2 96-041, Tcledyne llPSI Class 1 Piping Stress Analyses, Thermal htargin/ Low Power and Power Density Trips; October 28 - Penetrant Examination, llPSI Pump Testing SP2604A&Il Procedures, Steam Generator Replacement Project, Seismic Evaluation Report (SQUG),

OP2344A 480 Volt Load Center, Licensing Commitments; October 30 Corrective Actions, Containment Sump Level, Steam Generator Replacement Project, Exclusion of Faulty / defective Parts, Corrective Action Closcout; and Noverr.ber 4 PDCR 2 121 81, Reg Guide 1.97 Requirement on Nuclear Flux Measurements, Accumulator Tank Support Steel, ASME Section XI List of Repair / replacements. Tier 2 Accident Mitigation System Review RAl's.

Please call me at (610) 855 2366 if you have any questions.

Sincerely, a

, an m

Daniel L. Curry 7

Parsons ICAVP Project Director DLC:djv

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Attachments -

1.

Telephone Conference Notes from October 23,1997 i()Cf 2.

Telephone Conference Notes from October 28,1997

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3.

Telephone Conference Notes from October 30,1997 4.

Telephone Conference Notes from November 4,1997 cc:

E. Imbro (2)- USNRC J. Fougere - NNECo 11 Eichenholz-USNRC Rep. Terry Concannon NEAC R. Laudenat NNECo Project Files

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A!MINISTi(ATIVE CONFERENCE NOTES October 23,1997 i

PNRPOSE:

Administrative telephone conferente with NNECo, NRC, NEAC and Parsons to I

discuss:

Containment Spray Calculation Applicability of Safety Guide 26 / Regulatory Guide 1.26 e

• Use and Control of SF Forms Specification Revision e

Jumper Desice Index No. 2 96 041 e

Tclodyne llPSI Class 1 Piping Stress Analyses Thermal Margin / Low Power and Power Density Trips Ni Calibration at Startup Date: _

October,231997 List of Attendees:

NNECo NRC NEAC Parsons Joe Fougere Manager, ICAVP Steve Wayne Reynolds Dobson Fnd Mattioli Supenisor, MP2 ICAVP Eric Blocher Ibb Borchert Supenisor.Tecimical Support Eng Mike Akins Rich Ewing Supervisor. Design Engineering John Strange Willic Williams Enginect. CMP Jaun Cajigas Lou Chiarli.la Engineer. MP2 Projects Bob Moyer Marty Van llaltern Supervisor.Thermalliyd aulics Richard Moyer Dave llajumpaa Engineer.Thermalllydraulics Samir Serhan Mike Gancart Engineer ABII/CE Paul Schmitzer Robert McIleth Engineer CMP Chi Wu Engineer MP2 Fuels

l. Contninment Spray IMGGRQUNQ :ABB Calculation no. 006-AS95 C-017 documents the containment LOCA and MSLH analyses of record The MSLB analysis uses the SGN 111 brailt in containment spray emciency per a CCT(32) input of zero. The LOCA analysis spray data from ;he LOCA CONTRANS input is used in a similar fashion. We received SGN ill Figure 6 which apparently represents the code's built in spray emeiency data.

J QUliSllDR llow is the SUN ill and CONTRANS analysis spray emeiency data related to the MP 2 containment spray nozzle design and design flow rate? Tie design flow rate for the system is 1300 gpm per pump and 700 microns drop diameter at this flow rate.

Discussion

• - Topical Report CENPD 140 A explains that the spray emeiency is based on a 1000 micron drop diameter, and is supported by actual test data. Millstone's 700 micron drop diameter would be conservative.

• Millstone confirmed that they are currently redoing the LOCA and Main steam line break analysis. These should be complete in December.

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A DMINISTRATIVI: CONF t:RI'.NCI: NOT I:S October 23.1997 2, Safety Guide 26 / Regulatory Guide 1,26 (from 10!!6 conference)

Il6G GRQUND: It is not clear to what estent Safety Guide 26 / Regulatory Guide 126," Quality Group Classification end Standards"is applicable to Millstone Unit 2.

LH# 1350 From Pl 07 for AFW gives two different statements of compliance with Safety Guide 26.

'!he first reference is to Section 5.2 of the SER dated May 10,1974, from which it is concluded that Safety Guide 26 / Regulatory Guide 1.26 is applicable to Millstone Point Unit 2. The second is referenced to a response to question 4.7 of NU Amendment No.15 to License Application dated February 16,1973 from which it is concluded that only certain portions of the AFW System are in compliance with SG 26.

Although these items relate specifically to the design codes identified in SG 26 for the difTenent quality groups, it is not r. lear to what extent the guide as a whole is apphcable to Millstone Unit 2.

In a letter to the Commission dated September I,1972, " Request for Full Tenn Operating License", it is stated that 'NUSCO Quality Assurance Manualimplements S<cty Guide 26 as follows:

1.

Part I, Section 2A of the manual provides guidelines to the engineer (designer) for determina: ion oflevels in accordance with the Safety Guide.

2. Appendis 111 of the manual delineates the various levels as determined in Part I, Secdon 2A.

QUES 110N: What does the NUSCO Quality Assurance Manual cover in relation to the guide? Ilow are the Quality Group classi0 cations indicated for the systems and interfac of components, and what are the requiremeats of the interface between systems of different classi0 cations? Is this perfonned to any industry standards?

Discussinn Millstone is not committed to Reg, Guide 1.26 cxcept as stated in its position paper. FSAR Table 4.2-4 lists the quahty classes.

When looking at P&lD, one can tell the piping claes by pipe identification number. Speci0 cation ME 688. (sometimes referred to as the MS3 spec) identifies how to read and interpret the pipe number. For Mdistone, the design code class = the safety class. The P&lD does not contain any class boundary infonnation.

Hy looking in MEPL, one can detennine equipment classificadon. Specification ME444 addresses MEPL.

'lhe discussion regarding Safety Guide 26 in the 9/1/72 " Request for Full Tenn Operating License" letter, is only applicable to Millstone Unit 1, tu Unit 2.

The NUSCO Quality Assurance Manual makes no commitment regarding the safety guide.

3. SF Forms (from 10/16 conference) llAGGRQMML We have seen the following fonns t. sed to prepare PCDR's and PDCE's:

SF 305. All reusions. "PDCR Checklist" Procedure and Form.

SF 325 All revisions. " Notice of Design Change" Procedure and Foim.

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ACMINISTRATIVE CONFERENCE NOTES October 23,1997 f

t' SF 327 All revisions. " Plant Design Change Request". Procedure and Form.

SF 359 All revisions. "PDCR Evaluation". Procedure and Form.

QUESIJON: Were these SF forms part of an approved design control program? Was the use of these forms governed by a procedure? If so what were the procedures? Are these forms or siuitar forms currently in use? Please provide a history as to the use of these forms.

jiESPONSE: SF forms were part of NOP 3.03 and various Admin. Control Procedures, (ACP).

1he DCM manual uses similar fonns. NNECo will provide a document listing the fonns and the controlling procedures.

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4. Specification Revision in reviewing GRITS for selection of a specification sample, we have run across a number of examples where the speci6 cations are in some level of revision and there are no DCN's posted against the speci0 cation.

Examples are:

Specinctfon 25203 SP EE 101 SpeciGcation for Furnishing Construction Senices to i

Millstone 2 Revision 4; Speci0 cation 25203 SP EE 105 Speci6 cation for Spare Reactor Coolant Pump Motor Millstone Unit 2 Revision 7 Speci0 cation 25203 SP EE-095 Speci6 cation for Electrical Installation Requirements for the Shutdown Cooling isolation Valve in light of paragraph 3, of NGP 5.11, how can a speciScation be revised without a DCN7 RESPONSE: NGP 5.1I was canceled in 6/95. Chapter 6 of the Design Control Manual,(DCM) allows a specincation to be revised without using a DCN.

5. Jumper Device indes No. 2 96-041 DAGOROUND: Jumper Device Index No. 2 96-04I consists of twojumpers installed to allow the Enclosure Building Filtration lleaters to operate with the Containment flydrogen Purge valves open or close. (Nonnally any one of the four Purge valves opening will turn olTboth heaters.1his is to prevent an explosion when exhausting hydrogen.)

The Technical Evaluation for the Bypass jumper states "A common mode failure has been found l

with the connection between the four Ilydrogen Purge valves and the Enclosure Building Filtration ileaters. To solve the conunon mode failurefor the short term, (emphasis added) during modes 3,4,5 and 6, the Enclosum Building Filtration 11 eater will be on whenever fan F25.8 e F25B aic in senice.

QVliSllOE What is the common mode failure that the Jumper Bypass is solving during the modes stated above?

The Jumper Devices were installed in April,1996. What action has been takcn to provide a permanent solution?

. Discussion Thejumper devise was approved in response to a LER written when it was discovered that if a purge valve failed to close, and then a I.OCA occurred, the Enclosure Building Filtration licaters would not PAGE 3

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ADMINISTRATIVE CONFERENCE NOTES j

Octotter 23,1997 work. *lhe heders are needed post LOCA to protect the charcoal filters. A permanent modi 0 cation has been approved to correct this situation, M?. 97 013. Procedure OP 2313C was afketed by this change.

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6. Teledyne llPSI Class ! Piping Stress Analyses Regarding the Tcledyne llPSI Class I piping stress analyses and Stress Repons that we have received i

to date, most of the calculations src of 1975 vintage. No Calculation Change Notices (CCN's) have i

been supplied with these calculations. Ilowever, in at least one case (reference RAl 0665), there is an indication that other analyses exist for this piping.

ilased on the above, are the supplied Tcledyne calculations the latest calculations for the piping system, i

or do others exist that may eithee augment or supersede these calculations?

Discussion NNECo is sending Parsons any CCN written for calculations we request. Ilowever, CCN use started in the late 1980/carly 1990 time frame.

• For the ital referenced. NNECo reviewed the PDCR and found that a memo was used tojustify the reme val of the suppon. NNECo is investigating if a CR should be written.

Prior to the CCN, nothing tracked a calculation change. A memo could hr.ve been used tojustify a change, and it may NOT have been reficcted in the calculation of record.

In response to a related question, NNECo will look up the PA 81-005 file to find information on locked t p snubbers. Parsons will prepare an RAI to request this information.

7, Thermal Margin / Low Power and Power Density Trips Many accident aralysis use a program to calculate the Thermal Margin / Low i:ower and Power Density Trip points. These values are then used in the analysis to consider the timing of the accident and its severity. The formulas used for these trip noint calculations are not referencoj in the analysis. For the accident antlysis to be valid, the formula used must be the same as used in the plant calculators, is there docunientation available which can be usal to verify that the same formulas were used?

l Dis cussion To clarify the above question, lhe accident analysis calculations determine the conditions at which a plant trip should occur, llow does Millstone verify that the plant electronics will trip the plant at the same conditions Thermil Margin / Low Power and Power Density Tiip point logic is hard wired. IAC procedures setup and validate that the plant will trip as required by the accident analysis.

8. Ni Calibration at Startup Procedure EN21022 paragraph 1.7A-6 requires verification of excore response when between 2% and 5% power but does not state how this is done. NU's response to SOER 90 3 stated that Reactor Engineerir g uses core differential temperature to verify the Ni calibration on startup. Ilow is the verification required by EN21022 performed and what controlling document insures the methodology j

includes this core dilTerential temperature check?

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A*,MINISTRATIVE CONFERENCE NOTES f

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Discussion

• The step in procedure EN21022 paragraph 1.7.4-b is a result of a 1984 incident where leads on an excore detcetor were reversed. VeriGeation of the response simply ensures that the excore detector is working properly.

l It.itially, at the start of a new fuel cycle, the NI is setup based on evaluation of excore current readings vessus Dux levels during the previous cycle. Weighting factors adjust for the new cycle

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based on predicted fuel aux density from the reactor core analysis. At 20% power a calometeric j

measurement it, performed to check plant power level and the NI is adjusted as needed. This is repeated at 30% power level.

• - Core AT is an input to the reactor protection system. At low power levels AT is sinall so it is not a l

Very accurate measure of power.

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A!MINISTRATIVE CONFERENCE NOTFS October 28,1997 PURPOSEt Administrative telephone conference with NNECo, NRC, NEAC and Parsons to discuss:

e Penetrant Esamination HPSI Pump Testing SP 2604A & il Procedures e

Steam Generator Replacement Project e

Seismic Esaluation Report (SQUG) e OP2344A. 480 Volt IAad Centers e

Millstone Unit 2 Licensing Commitments e

Date:

October,281997 List of Attendecs:

i NNECo NRC NEAC Parsons Joe Fougere Manager, ICAVP Steve Wayne Dobson Reynolds Fred Mattioli Supenisor, MP2 ICAVP Eric Blocher Steve Wainio Supenisor Design Engineering Greg Cranston Dill Price Supervisor. Design Engineering CliffMarks Willic Williams Engineer CMP Don Marks Sal Orelice Supervisor Ray Thomas Mike Mcdonald Engir..:t CMP Ken Gabel John llemis Engineer MP2 Technical Support Samir Serhan Rick Donner Supenisor MP2 Operations Larry Collier Prem Godha Engr.. Nuc Materials Engineering Scott Duplantis Engr Nuc Materials Engineering Charles Peterson Engr Nuc Materials Engineering ining Tsang Engineer - CMP

l. Penetrant Examination lias a high temperature (> l25 deg. F) and/or a low temperature ( < 60 deg. F) qualification test been pc formed for the liquid penetrant procedures NU LP-1 and NU-LM as required by ASME Code Section V Article 6 Subarticle T-680, or has relief been granted from this requirement either through an approved code case or some other regulatory document?

RESPONSE: The procedures were qualified to the low temperature requirement by following a practice used by other companies, i.e. the metal is cooled to less than 60 deg. F. When the liquid penctrant is whied, it has such a thin film that its temperature immediately drops and matches the metal.

2, llPSI Pump Testing SP 2604A & H Procedures Are the flow rate, pressure and vibration measuring instruments analog or digital in construction e

detection and indication?

What is the maximum range of each of these instruments?

e-e.

What are the incremental changes for each of the instruments at their test measurement points?

is a Post Test calibration routinely performed on each of these instruments or is the calibration e

performed on a specified periodic basis?

ifinstruments are found to be out of calibration, what is the course of action?

e Discussion

• Min, flow instruments are digital reading 0.lgal. and calibrated i 2% liigh flow is analog PAGE 1

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A;MINISTRATIVE CONFERENCE NOTES October 281,1997 with a digital readout,0 300gpm range, i 2% full scale. Vibration instruments arc analog with digital readout,0 30 Khz range,15%.

Flow and pressure instmments are calibrated every 18 months. Vibration instruments are calibrated annually.

• If an instrument is found to be out of calibration, evaluations are donc to determine; l) the validity of the measurements taken and 2) the operability of the equipment. Appropriate actions are taken based on the results of the evaluations.

3. Steam Generator Replacement Project a) llow was the impact on t'ic feed water system supply lines evaluated and documented with respect j

to increased vessel mass, new steam generator CG, and changes in thermal as well as scismic design inputs?

Reference:

piping line number 18" EBB-6;- pipe stress analyses problem numbers 21 and 22.

b) llow was the impact on the instrument tubing and ttibing supports including location (orientation and c!cvation), rcrouting required, fitting changcouts evaluated and documented? Ilow were any performed modifications documented (i.e., DCN's) and justined from a calculation standpoint.

What document (s) identifies actual elevations for all " Instrument Nor21cs" and what documentation (i.e., DCN's) was used tojustify these locations for a setpoint calculation (s)?

Discussion The NSSS was evaluated including new CG, thermal characteristics, and motions versus e

clevations. This information was contained in a single transmittal to NNECo which was provided to Fluor Daniel for revision to the feedwater analysis and for the tubing work, if Parsons identifies a tubing DCN ofinterest, NNECo will track down the calculation that e

supperts the DCN.

c) llow were changes on thermal movements for SG nor21es evaluated, and where can these design inputs bc found?

c) What was the drawing control process? Specifically what is the relationship of the Fluor Daniel drawings to the NNECo drawings (i.e.,86242 284081022 vs. 25203 284081022).

Discussion For this large project, the original drawings were copied, the normal 25203 id number was e

replaced with the pA#, (86242) and these were provided to the A/E for their use. After thejob was complete and ready for as building, the PA numbered drawings were reconciled with the record drawi,gs and incorporated into a revision of the 25203 series drawings. During thejob, the PA# drawing resisions were controlled via rev. a, b,c, etc. All of the PA# DCNs have been incorporated into the record drawings.

Currently Raytheon is doing a drawing update backlog reduction cfTort. For Unit 2 there are a

approt 2500 drawings vith outstanding DCNs,(none of which are Ops Critical). Of these, approm 700 need to be need to be complete piior to restart.

Prior to 1992, NNECo could change a drawing with out a DCN. They would change the drawing directly with a revision instead ofissuing a change notiec against a drawing, which would later be incorporated into a drawing revision.

d) We need clarification on where is the cut point line location on drawing 25203 29145 sht. 296 for the steam generator replacement component (lower assembly).

O What is the relationship between Fluor Daniel specifications NU 7 K012/0.830100 70200/0 (instrumentation) and NU $0-5002/0; 830100 50955/0 (root valves) and the respective Bechtel Specs 7604 MS-66 and 7604 MS-647 Which one is the design basis for the steam generator..

- replacement? BES_PONSE: NNECo will provide info during the 10/30/97 Conference.

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AP MINISTRATIVE CONFERENCE NOTES October 241,1997 i

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4. Seismic Evaluation Report (SQUG)

The Seismic Evaluation Report (SQUG) walkdown identified outlicts for equipment quahrication. The response to these outliers was recommended hardware changes listed in the Seismic Evaluation Report.

What change centrol documents ensure that these hardware changts are implemented?

Discussion Closure of 87-02 will be donc via a letter to the NRC when everything is complete. The actual i

modifications are accomplished via the Design Control Manual.

Relocation of a relay is still pending. There was one relay that NNECo thought was a problem, but it was later determined that the relay was class 1E and qualified.

NNECo can cross reference the 87-02 closure to specific modifications

5. OP2344A,480 Volt Load Centers Regarding OP2344A,480 Volt Load Centers, Section 5.17 5.29 flow often are the mentioned jumpers used?

e Does the Temporary Power to the RWST IIcat Trace maintain operability?

e RilSi10 HSE: The jumpers are used once per refueling outage to de-energize a load center for maint, while still providing power to items which need power to support the outage. The jumpers are provided for in the procedure because they are used regularly. *lhe RWST lleat Trace operability depends on temperature, not electrical power.

6. Millstone Unit 2 Licensing Commitments in several different documents we have noticed terminology which sounds like there is a data base for i

Millstone licensing commitments. For example, UIR 2740 contains statements like, " Commitment record XXXX" states.... Does a commitment database exist? Is this terminology referring to I

snmething more than LIST or the PI-06 evaluations?

BilSPONSE: Yes. i. commitment database exist. It is a ACCESS database created for the cmp, pI-06 populated this database. parsons has a copy of the database. Its file name is CMpMASTER.DNB.

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ADMINISTRATIVE CONFERENCE NOTES Octeher 30,1997 PURPOSE:

Adadnistrative telephone conference with NNECo, NRC, NEAC and Parsons to discuss:

• Corrective Actions Containment Samp Level e

Steam Generator Replacement Project (follow-up from 10/28) e Esclusion of Faulty / defective Parts e

e Corrective Action Closcout I,ist of Attendus:

NNECo NRC NEAC Parsons Joe Fougerc Manager, ICAVP Steve keynahls Mike Akins Fred Mattioli Supervisor MP2 ICAVP Ikic illoches Dill Price Supervisor Design Greg Crancon Engineering Willie Williams Engineer CMP samir Serhan John Hemis Engineer MP2 Technical Ken Unbel Support John Resclar Manager MP2 Corrective Ray Wmas Actions Mark Suprenant Manager Procure Dan WoodJell Enginecting Lou Chiariria Engineer MP2 Projects Wayne ChommansLi Philliiggins Engineer MP2 Design M.vL litgerald Engineering Gary Komosky Engineer MP2 ICAVP

l. Corrective Actions For UIRs that were closed as being picked up by PI 07 review by NNECo, we can not find any reference in P! 07 where numerous UIRs were addressed. Specifically UIR's 1052,1072,1104,1286,1398, and 1616. We espect that other UIR'S fall into this same category, llow did NNECo documented closure of UIRs as part of Pl.077 During preliminary review of the summaries of closed CR's on the s<:lected Tier i systems we found CR M2 97 0900 was closed out to a UIR. This CR referenced UIR 2521 as correcting the issue. Does the NNECo process allow closure of ACR's back to UIR's without issuing a specific corrective action under and AR.

Discussion i

PI 14 covers the UIR process used by NNECo. Some of the UIRs were written as informational UIRs and were closed because no corrective action was necessary. !! corrective action was necessary, then a CR was initiated against the UIR. An AR could have been written that does not attach to a CR. AR can be closed in a CR if the AR is a long term monitoring issue, or if the department head closes it.

2, Containment Sump Level NNECo indicated that Sp 21136 is the procalute which directs personnel to monitor or measure that piping is maintained with borated water filled to an elevation of(-) 24' or higher in order to prevent pressure locking of Containment Sump Isolation Valves,2 CS 16 l A and 2-CS 16.lD. What documentation would identify the required frequency for this inspection, and that the inspection has actually been performed? We have looked in PMMS for AW0s associated with these valves, but have been unable to kwate anything that indcates that this work was perfonned PAGE 1

ADMINISTRATIVE CONFERENCE NOTES October 30,1997 i

Discussion SP 21136 has been done on July 21,25 '96; October 1I, '96 and Jan 10 '97. Afler that time the plant was shut down. The valves are tested per ash 1E {XI per IWV 34.11 on a quarterly basis.

There is a grace period of approximately 22 days (25% on TS) associated with this. This prowdure uses a temporary hose that delivers 15 18 gpm to fill the sump. De procedure will be deleted when a modification installs thermal locking hardware. He modification is currently in developmer;t and is a startup item.

QUliSIlOM Concerning IIR 95-002-00, Anchor Darling performed an analysis and determined that the maximum pressure lock that containment sump isolation valves 2-CS-16.l A & B could overcome and still open is approximately 150 psi.

What Continment Side, RWST Side, and Valve Bonnet pressures were assumed for this analysis?

What were the method and assumptions used to calculate valve thrust?

e What was the value of the reduced voltage factor that was used in the analysis?

e Discussion There is a letter from Anchor Darling pertaining to these valves. It is based on 150 psid for the Bonnet, with 0 psid up snd down-stream of the discs. Here are two calculations for the valves.

One for the operability of determination for the valves and one for the design conditions of the valves. He operability calculation does not use degraded voltabe, while the design calculation does use degraded vol age.

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3. Steam Generator Replacement Project (follow up from 10/28) d) We need clarification on where is the cut point line location on drawing 25203 29145 sht.

296 for the steam generator replacement component (lower assembly). Specifically we are looking for the distance between the upper and lower tap for the narrow range steam generator level. instrun entation. Also, does DCN h12 5 546 91 address the new of old steam generator?

Discussion The drawing that contains the dimensional information for the steam generator is 28408 sheet 1022. This is currently in design engineering foi incorporation of DCNs. DCN-m2-5-546 91 will b in the current update.

Elevations were developed based on containment elevations measured to the base reference point on the floor. RAI was issued to request the drawing and to receive a response to the question:

"What temperatures were the dimensions based on?"

f) What is the relationship between Fluor Daniel specifications NU 7 K012/0:830100 70200/0 (instrumentation) and NU 50 5002/0. 830100 50955/0 (root valves) and the respective Bechtel Specs 7604 hts 66 and 7604 hts-64? Which one is the de,i n basis 3

for the steam generator replacement?

Discussion Installation was based on a Bechtel Specification 7604 h1S w and 7604 h1S 64 prepared specifically for htillstone 2, and it contains more information than was used. (Note: this was requested under RAl 739)

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ADMINISTRATIVE CONFERENCE NOTES October 30,1997

4. Earlusion o Faulty / defective Parts r

Several regulatory documents such as NRC Ilulletins, LER's etc. pertain to defective components or subcomponents. In many cases Northeast was asked to verify whether or not the subject components or sabcomponents were installed at Millstone 2, and if so in what capacity, i c. safety related or non safety related. Most of the documents reviewed were from the mid 80's time frame or earlier.

QUESTION: What barriers doca Northeast have in place to ensure the exclusion of these items? Do you have a document uhich specifically mentions these faulty / defective parts and is it procedurally referenecd in the procurement process? Is this issue addressed in the case of commercial grade components? (Noic: we will ask for a copy of documentation that addresses this issue)

Discussion The plant used the NODIL (Nuclear Opera' ions Defective items List) since the mid-80's for this.

The NODIL is updated annually, it was updated in Jan 97. All the fields are "Q"and controlled by a select few. NGP 6.02 points to the use of MpM 3.00 (basic MIMS procedure) and this points to MPM 3.05, the Admin. procedure.

5. Corrective Action Closcout ilACKGROUND:. A review of two closed ACR's does not show traceability or linkage to associated AR's. A closed ACR could be reviewed and found acceptable without knowing if all AR's associated with that ACR were captured and properly closed. For Open ACR's that we will review to detennine if we agree that it can be Open for restart, vce have not found an electronic way to link associated AR's to ensure all relevant documents have been re' iewed to make the correct decision.

For example, ACR 8490 is shown as closed or complete. The AR numbers referenced on the cover are 96002343-01 and 96002343-02. A different document indicates that AR's 96002343-03 and 9700198517 a also applicable to ACR 8490 and would have to be closed to allow the ACR to go to complete. Iloweecr the later two AR's are not referenced by number anywhere in the ACR. Also, none of the A P.'s are included as part of the closure package sent to Parsons, i

l A similar situation exists with ACR M2 96-0348 and AR's 96030546-02,96030546-03 and 46030546-01. The first two AR's are referenced on the cover sheet of the ACR but the third is not referenced at all. None of the AR's are included in the closure package.

The Utility response to RAl-00490-10/8/97 provided a current list of ACR's/CR's and their status, with a summary. Ilowever, the associated AR's are not listed.

QUESTION: Is there a Database that capteres and ties ACR's/CR's to associated UIR's and AR's so that completion or e...yability for deferral can be detennined? Ilow does NNECo track ARs associated with and ACR/CR7 Discussion in the passport system there is the abihty to tie the ACRs and CRs and ARs together. Modes associated with the ACR are identified at the action level. It will be necessary to review all the

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acts associated with the ACR in order to determine the closure of the ACR and identifying ifit must be ccmpleted prior to restart. It is possible to sort the actions using the keyword "mid 13" PAGE 3

4 ADMINISTRATlVE CONFERENCE NOTES October M,1997 AITTS has a completc record of corrective actions, note tint it does not contain the full document associated with the action, but does reference it.

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ALMINISTRATIVE CONFERENCE NOTES Nm ember 4,1997 PURPOSE:

Administrative telephone conference with NNECo, NRC, NEAC and Parsons to discuss:

e PDCR 212181 Reg Guide 1.97 Requirement on Nuclear 11ux Measuren.ents.

Accumulator Tank Support Steel ASME Section XI 1.ist of R< pair / replacements e

Tier 2 Accident Mitigation System Review RAl's e

List of Attendecs:

j NNECo NHC NEAC Parsons Joe Fougere Marunger, ICAVP Steve Wape Ikhnon Reynolds Fred Mattioll Supenisor, MP2 ICAVP Sarmt Serhan Sleve Wainio Supenisor Design Engineering Rich Glaviano Dob llorchert Supenhor. Technical Support Dan Cardinale Eng Dave IWumpaa Engineer -Thermal llydraulics larry Collier Jim Diluca Design Engineering 11ob Moyer Paul Wagner Manager Desiga Kalvin Anglin Design Engineering 1.loyd !!aird Technical Support Eng Ken Fox Supenisor Designlingineering Mike Kai Supenisor Safety Analysis

1. PDCR 212181 PDCR 212181, titled " Redesign of Pipe Support 401106" proposes to modify this sepport to meet IEll 79-02 an,17914 criteria. Ilowever, within the PDCR, the Safety Evaluation, Design inputs, work authorization (PA 79 176) etc. appear to adhess all 79-02/7914 rework.

Does this mod cover all 79-02/7314 rework, all llPSI 79-02/7914 rework, orjust 401106 rework?

KliSPQNSji: Just the 401106 rework ifjust 401106, is this the only llPSI support that required rework for 79-02/79 147 e

if additional llPSI supports required rework, which ones and which mod!(mod's) e addressed the work?

R[iSPQNSji: 401106 is not the only llPSI support that required rework for 79-02/7914. In response to RAI 491, Scheduled for I t/13/97, other rework for 79-02/79-14 will be provided.

When could we get the support design cale for 401106, request by RAl-Ol97, dated 8 5-e 977 R[iSP_QNSII: 401106 is currently being reanalyzed. It is anticipated that as part of some RWST work,401106 will be removed.

2. Reg Guide 1.97 Requirement on Nuclear ilus Measurements.

BACKGROUND: Reg Guide 1.97 requires the display of wide range nuclear flux measurements for monitoring attainment of sh 2tdown conditions and the potential for a recriticality accident.

He initial design proposed by NNECo (1114034,3/2/92) utilized all four wide range channels as PAGE 1 e

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ADM INISTRAllVI: CONFl:Rl:NCI: NOTI:S Nm ember 4,1999

, accident monitoring instrumentation. The current design (SP Mp 1?E-001212/2/96) takes credit for only two channels (A & D). Ilowever, as-built design includes four wide range logarithmic neutron monitoring channels spaced around the core to measure power and rate of change of power in startup and interm(xliate pcme levels. 'these signals are provided to the operator on the main control boards.

QUliSTION:

a) lias a, analysis been performed that addresses the abdity of the two selected channels to detect approaches to recriticality under conditions which may involve substantial flux tilting (Cl!A drop or ejection). If yes, please identify and provide the document.

b) 3ince all four channels are displayed on the main control board, is there a technical reason uhy only two of the channels are designated as R.G. l.97 accident monitoring instrumentation?

Discussion No analysis has been done that addresses the ability of the two selected channels to detect approaches to recriticality under conditions which may involve substantial flux tilting. For the examples of CEA drop or ejection there would be reactor shutdown with no return to criticality unless there were multiple failures, so such an analysis is not needed.

A letter to the NRC dated i1/7/95 changed the RG l.97 designated channels from 4 to 2. Only the A & D channels are EQ qualined due to a problem with the H & C channel cable mineral insulation.

IIACKGROUND. Redundancy requisements of R G.1.97 are based on the premise that both of the redundant channels are monitoring the same variable, and that the variable will have essentially the same value at both points of measubment. R G.1.97 states that "...(beyond redundancy)... It is important that the number of points of measurement be sufficient to adequately indicate the variable value.. " This criteria is typically applied to containment temperature, but applies to neutron Hux measurements as well.

QUESTION: lias NNECo analyzed the ability of the system to provide correct information (power level and rate of change) to the operator under conditions of single failure of one of the two channels and substantial flux tilting within the core?

Discusslun Na analysis has been Jone that addresses this area. Nonnally, the plant will have 4 channels to detect recriticality, except for situations that result in a harsh environment

3. Accumulator Tank Support Steel With regard to original equipment, safety related accumulator tanks, such as T123 A&R and Tl24A&D shown on p&lD 25203-26015 Sheet 1(J 7), and T121, T122 Tl33 & T134 shown on P&lD 25203-26028 Sheet 3(G-4), we are trying to determine what document provides the criteria / instructions for mounting these tanks to their support steel. Spec. 7604 MS-66 Rev 7,

" Design Guide For Seismic Class I instrument Tubing Installation " states in Section 3.7, "If accumulator tanks and check valves are required on control valses, they shall be installed in accordance with this document or Scismic Class I Criteria". A review of MS-66 indicates no instructions for accumulator tank installation. Therefore, what is the Seismic Class I Cnteria referred to and is there any other documents that provides the criteria / instructions for mounting these tanks to their support steel?

RUSMN_SE: 1he accumulator tanks were purchased to Spec. MS 226. Bechtel calculations 4

& 6 designed the tank supports.

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ADMINISTRATIVE CONFERENCE NOTES November 4,1997

4. ASME Section XI List of Repair / replacements Reference RAI 426 response, list of ASME Section XI Repair and Replacements. One of the column labels is "Sent to File". What does this indicate? Many of the items completed in the j

1992, '93, and '94 time frame have no date listed for "sent to file". The purpose for this question is to detennine what documentation is available for the ICAVP resiew.

BESPONE: Any item li ted with an AWO that has been completed is available for the ICAVP resiew. "Sent to file" has no relation to the information being sent to nuclear records.

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5. Tier 2 Accident Miligation System Review RAl's RAl's 423 & 427 were submitted to gather design information on the Si Tanks and the Shutdown Cooling Systems. Additional RAl's will be subinitted to obtain design information on the remainder of the Tier 2 Accident Mitigation systems. Based on Parsons review of the infermation submitted in response to RAl's 423 and 427, we propose the following approach:

1) Parsons will submit specific RAl's for each of the remaining systems. Each RAI will identify a specific parameter (eg, flow) and request parameter specific information that can be determined from sources available to Parsons This information should be prosided to Parsons within 2 weeks of request.

2) In addition, we request that the NNECo System Engineer review thn RAI and; a identify additional parameter specific information relevant to the system,

b. identify calculations listed on the RAI which have been superseded or that are being revised.

3) is there a limit to the number of documents that can be requested on an RAl? If we need 50 cales, should we make that into multiple RAl's?

Discussion The above approach is acceptable. One H M is desirable, there is no limit to the number of documents that can be requestal on a RAl.

s

4) Severa! FSAR Chapter 14 analyses are being updated by NNECo. MSLB is an example.

Parsons is performing the Tier-2 rniew using the current analyses. We will review the revised analyses near thc end of the Tier 2 review and address any changes that have been incorporated into the updated analyses. We would like to obtain the updated analyses as soon as possible (by early December, if possible). Which analyses are being updated, and can we get copics of the updated analyses by early December?

Discussion NNECo is currently preparing inputs for the Chapter 14 analyses and rniewing the schedule. It is unlikely that the updated anal ses will be done by early December.

3 The following are being reanalped:

Chapter 14.1. Excess Load. Main Steam Line Break Chapter 14.2 Loss of External Load, MSIV Closure l

Chapter 14.6 Small LOCA, Large I OCA Chapter 14.8. Main line Break / Containment Analysis Startup Rod Withdrawal Accident

• NNECo will send Parsons the schedule w hen it is available l

NNECo will send Parsons the rnised analysis inputs when they are available. Parsons can start their rniew withjust the inputs before the re-anas3 sis is completed.

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