ML20217P039

Transmits Summaries of Telcons Between Parsons Power Group, Inc,Nrc,Nneco & Neac on Noted Dates Re Independent Corrective Action Verification Program

ML20217P039
Person / Time
Site:	Millstone
Issue date:	04/30/1998
From:	Curry D External (Affiliation Not Assigned)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUM2-PPNR-1419, NUDOCS 9805060046
Download: ML20217P039 (12)

Text

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9 PARSONS Daniel L Curry, vice President. rextear services Parsons Energy & Chemicals Group loc.

2675 Morgantown Road

• Reading, Pennsylvania 19607 + (610) 855-2366

• Fax: (610) 855-2002 April 30,1998 Docket No. 50-336 Parsons NUM2-PPNR-1419-L U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 Millstone Nuclear Power Station Unit No. 2 Independent Corrective Action Verification Program (ICAVP)

Gentlemen:

This letter transmits summaries of telephone conferences between Parsons Power Group Inc., the U. S.

Nuclear Regulatory Commission, NNECo and NEAC on March 31 (Rev.1), April 2, April 7, April 9 and April 14,1998.

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

Sincerely, Daniel L. Curry Parsons ICAVP Project Director

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DLC:djv Attachments 1.

Telephone Conference Notes from March 31,1998 (Rev.1) 2.

Telephone Conference Notes from April,2,1998 3.

Telephone Conference Notes from April,7,1998 4.

Telephone Conference Notes from April 9,1998 fff)O /

5.

Telephone Conference Notes from April 14,1998 i

cc:

E. Imbro (2) - USNRC J. Fougere - NNECo H. Eichenholz - USNRC Rep. Terry Concannon MEAC R. Laudenat - NNECo Project Files 9905060046'900430 3

PDR ADOCK 05000336 P

PDR PPNR1419. doc

ADMINISTRATIVE CONFERENCE NOTES Mcrch 31,1998 DATE:

3/31/98 PURPOSE:

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

1.

Alternate Shutdown Panel 2.

Surge Suppressers for MOV SV 4188 3.

Status of EWR's included in the Tier 3 review sample 4.

NNECo response to DR-0160

5. Specification Revisions 6.

Discussion of DR-127 7.

Discussion of DR-il9 LIST OF ATTENDEES:

NNECo NRC NEAC Parsons Joe Fougere Steve Reynolds Wayne Dobson Fred Mattioli Don Marks Ken Fox Jim Giova Ken Moore Jim Collins a

Greg Tardif Raylhomas John Lockerby Kent Russell Dan Hunley Andy O'Connor Dan Van Duyne Dale Pruitt George Pitman Ken Mayers Cris Cristallo Larry Wigley Dave Bajumpa Bob Steirenetz Mike Akins Joe Groncki 1.

Topic: Alternate Shutdown Panel (Andy O'Connor)

Background:

N/A Questions:

a) Has the Unit 2 Alternate Shutdown Panel and its isolation transfer switch capability ever been functionally tested either during the original startup or since that time?

b) If so where would the documentation be found?

Response

Documentation cot:Id not befound regarding the testing ofthe Alternate shutdown Panel (C-21).

Individual components do have surveillance procedures in place See Tech Spec Table 4-3-6:

Comrument Procedure Channel Check SP-2619 E Channel Cobbration (Reactor ColdLeg Temperature)

No surveillance procedure (Note: a CR has been generated, a number to be supplied.)

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Revision 1 PAGE 1 l

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ADMINISTRATIVE CONFERENCE NOTES March 31,1993 Response Continued:

Comr>ortent Procedure Pressuri:er Pressure Low range SP-2402 J High range SP-2402 B Pressuri:er Level SP-2402 E Steam Generator Level SP-2402 D Steam Generator Pressure SP-2402 C 2.

Topic: Surge Suppressers for MOV SV 4188. Follow on questions to 3/26/97 Conference (Dale Pruitt)

Background:

PDCR 2-89-107 (Surge Suppressers for MOV SV 4188) identified that during a walkdown surge suppressers were missing from SV 4188M. The modification evaluated surge suppressers to be installed. The modification package received does not indicate the suppressers were installed, however they still appear on elementary 32020, Sheet 49. The modification data base indicates the modification was canceled.

A search of the PMMS Database did not locate a work request that addressed the condition.

Per the conference of 3-26-97 three of the five suppressers were found in the 6 eld Questions:

a) What qualified the installed suppressers?

Response: Drawing 800256D b) What installed the suppressers since no work document could be found?

Response: PDCR 2-140-81, CR 98-0855 was writtenfor the missing suppressers.

Note: the PDCR and associatedAll'O will be requested on an RAl.

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3.

Topic: Status of EWR's included in the Tier 3 review sample. ( Bob Steinmetz )

Background:

After the review of the majority of EWR's selected for the Tier 3 review, we can not determine the open / closed status and the design change mechanism ( i.e.; MMOD, DCR, DCN etc. ) for three remaining EWR's.

Questions: Please provide status and design change documentation citations for:

a) EWR # 2 97-007, Thenno-Lag Fire Barrier Response: EliR is open andno design change is assigned Revision 1 PAGE 2 l

ADMINISTRATIVE CONFERENCE NOTES March 31,1998 b) EWR # 2-96-156, Condensate Min. Flow Recire Response: EliR is open and DCR # A12-97-520 was assigned.

c) EWR # 2-92-172, RBCCW Rupture Dise ( Note: RAl-586 stated this is within a previously submitted RAI-539. The design change documentation can not be located within the RAI response. Please provide additional information.)

Response: Status ofEliR ll 2-92-172: RBCCll' rupture disc EliR is still open and DCR 11 Af2 018 was assigned. Per the results ofthe conference call, NNECo was to submit the DCR on the next day, however, this was not received. II'ayne will discuss this with NNECo during the status meeting on Tuesday (M7/98).

4.

Topic: NNECo response to DR-0160 (Jim Giova)

Background:

PDCR 2-45-82 changed the thennal margin / low pressure (TM/LP) Pre-trip setpoint to decreased the margin from 100 psia to 75 psia above the TM/LP trip setpoint for the Reactor Protection System. This change was accomplished by changing two resistors in the Core Protection Calculator No. I of the Reactor Protection System (RPS) from a 1000 ohm rating to a 750 ohm rating.

Our review of GRITS indicated that there were no change documents against the three vendor manuals. The ven'.'or manuals are:

VTM2-150-00' A " Bistable & Auxiliary Trip Unit", Rev. O, 12-04-87.

VTM2-150-008n," Reactor Protection System (Vol.1)' Rev. 01,2-11-86.

VTM2 150-009A, " Reactor Protection System (Vol. 2)" Rev. O,12-07-87.

NNECo's response to DR-0160 stated that Vendor Technical Manual (VTM) VTM2 150-007A was revised to show the change. It did not identify whether the Thermal Margin Pre-trip Offset tenn, the 75 psia, or the change in the resistor ohm rating was changed in the VTM. The response also stated that VTM's 150-008A and 009A do not contain information that requires updating.

A search using GRITS was performed to determine the change -ontrol document that was used to update VTM2-150-008A. GRITS indicates that there are no change control documents posted j

against the VTM.

Questions:

a) When and how was VTM2-150-007A updated to reflect this change?

Response: NNECo stated that they cannot currently locate the vehicle used to update 17Al-150-00&t. NNECo said they wouldforward additional information to support their disposition of DR-0160 regarding the change to ITAL-130-00&1.

Parsons asked how we can concur with NNECo 's response to DR-0160 that i TAis -150-007A and 009A were not impacted by the revision to the Thermal Afargin Pre-trip Ofset term or the change of two resistors

• ohm rating. Parsons suggested writing a RAI to obtain copies ofthe vendor technical manuals (1 TAf), ifthey are not too voluminous. NNEco stated that they are reluctant toforward the 1 TAis because they will be revised by the reconstitution eforts ofDC-16.

NNECo concluded by stating that they willprovide additionalinformation to support their disposition to DR-0160.

Revision 1 PAGE 3 l

ADMINISTRATIVE CONFERENCE NOTES Mirch 31,1998

5.,

Topic: Specification Revisions (Jim Collins, Wayne Dobson) from 3/24/98

Background:

During the Tier 3 review of specifications, several specifications were noted to be active in GRITS but the specification does not contain the latest technical requirements.

Examples of these specifications and a brief description of the condition are as follows:

i SP-GEE-48 R1 is a purchase specification containing the technical requirements for the e

design, fabrication, qualification testing, packaging, and delivery of electrical connector / cable assemblies. This revision issued in 1980 could not be used today since it does not contain the latest environmental qualification requirements.

SP-ME-220 R1 is a purchase specification for eight safety valves which was issued in 1980. Based on the review performed the specification should not be used to purchase valves today since it does not appear to include all the latest technical requirements.

Specification 25203-7604-M-506 is an original Bechtel specification for the installation of containment and safeguards ductwork. The DCN being reviewed made reference that a new access door is a medium pressure door in accordance with the specification. The specification does not contain requirements for medium pressure ductwork, nor can it be conclusively determined if the specification has ever been revised or evaluated to determine ifit contains the latest requirements.

We have not found any problems with the use of these specifications in the past, and we understand that Millstone uses specifications that contain original design requirements when assessing existing plant equipment.

Questions:

a) What current procedure or process prevents specifications that do not contain the latest tedmical requirements from being used for NON-MODIFICATION work activities, such a procuring replacement parts? [For example, does a procedure exist that requires specifications to be periodically reviewed and updated to include changing requirements or is there a requirement to identify or classify a specification so that it is not used until a review confirms it meets the latest requirements?

Response: NNECo responded that the NGPs associated with purchase orders, replacement item evaluations, and commercial grade dedication requires that the purchase request be reviewed by engineering to ensure the correct EQ, seismic, and other technical requirements are properly imposed. Examples in procedures NUC MPM 3.01, NGP 6.02, and NGP 6.12 were given. A review of NGP 6.02, Quality Material Requests and Quality Purchase Orders, confir.sd that respcmsibilities and requirements are contained in sections 3.3, 5.5, 5.6, 6.1.1.3, 6.1.1.6, 6.1.3.1, 6.1.3.3, and 6.1.3.4. A review ofNGP 6.12, Evaluation ofa ReplacementItem, confirmed that responsibilities and requirements are contained in sections 6.2.2, 6.3.1.8, 6.4.2.1, 6.4.2.2, and 6.4.2.4. Therefore, it is concluded that the procedural controls in place should prevent the improper use of a specification that does not contain the latest technical requirements.

6.

Topic: Discussion of DR-127 (NNECo Requested Topic)

Response: NNECo presented their revised resp <mse to DR-127.

Revision 1 PAGE 4 l

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l ADMINISTRATIVE CONFERENCE NOTES Merch 31,1998 7.

Topic: Discussion of DR-119 (NNECo Requested Topic)

Response: NNECo presented their revised response to DR-119.

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l Revision 1 PAGE5 l

ADMINISTRATIVE CONFERENCE NOTES April 2,1998 DATE:

4/2/98 PURPOSE:

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

1.

Containment Free Volume 2.

Compliance with NRC Regulatory Guide 1.11 3.

CST Inventory Requirements 4.

EBFS System Scismic Duct Supports 5.

Valves 2-MS-432A and 2-MS-432B 6.

PDCE MP2-89-014 7.

Main Feedwater System Performance During MSLB Event 8.

Review of Engineering Specifications SP-EE-292 and SP-EE-105 9.

PDCR 2-231-76

10. DR-313

11. Containment Gas and Particulate Radiation Monitor Sampling Nozzics

12. Containment liigh Range Radiation Monitor Location and Installation Detail

13. Re-addressing Topic #1 of 3/31/98 (Note: not on original agenda)

LIST OF ATTENDEES:

NNECo NRC NEAC Parsons Joe Fougere John Nakoski Wayne Dobson Fred Mattioli Don Marks Greg Tardiff Bob Moyer Dave Bajumpaa Candace Segar George Pitman Richard Boyd Dan Sipple 1,ouis Mscichowski Clark Maxson David Lengel Farid Elsabec Gordon Chen Ken Fox Kent Russell Mike Champagne Rcger Mauchline Mike Smaga Jim Collins John Kapinos John Wilkens Andy O'Connor Mike Akins Dan Wooddell Iarry Wigley 1.

Topic: Containment Free Volume (David Lengel)

Background:

Bechtel Plant Data Book for MP2, section 6.1.1, identifics the containment free volume as 1,919,800 cubic fect. This value is also listed in section 3.1.1 of the Containment System DBD.

FSAR Table 14.6.5.1 3 identifies the maximum containment net free volume as 1,938,000 cubic feet.

FSAR section 5.2.1 identifies the containment net free volume as 1,899,000 cubic feet.

Bechtel memo MB-4742, Containment Pressure Calculation for ECCS, 9/30/74, identifies the free volume as 1,937,733 cubic feet. This value had been increased by 2%, the accuracy factor of the calculation, to arrive at a maximum containment net free volume. Removing this 2% factor yields 1,899,738 cubic feet as the calculated value which is very close to the 1,899,000 cubic feet value being identified as the minimum containment net free volume in FSAR section 5.2.1 PAGE1

ADMINISTRATIVE CONFERENCE NOTES April 2,1998 Per Bechtel memo MB-5235, Volume Verification Test, 8/27/75, the measured containment free volume was 1,919,800 cubic feet. This value was corrected, based on equipment which was vented during the test but would normally be isolated, and a corrected measured free volume of 1,909,000 cubic feet was determined.

All of the references listed above were created in the 1970's. Modifications within the containment have occurred between the time the free volume was determined and the present which could have reduced the containment free volume. Modifications, such as SG Replacement, can have an immediate impact on the free volume while other modifications, such as addition of piping restraints or addition of scaffolding storage boxes, may individually have minor impact but cumulatively, could represent an impact which must be considered. Existing margin in the containment free volume varies depending upon which free volume number is chosen but appears to be as little as 738 cub!; feet which could easily be exceeded based on modification work over 20 years.

Procedure EN 2l065, Rev.1, Containment Mass Trackmg, is used to determine changes to the net free volume of the Containment (steps 1.2,4.3.6). However, it does not appear that this process has resulted in any changes to the containment free volume as the current references remain the 1970 vintage documents.

Questions:

a) What value is currently being used as the containment free volume?

Response

Alinimum Containment Free l'olume = 1,899,000 cubicfeet Alaximum Containment Free l'olume = 1,938,000 cubicfeet b) What document contains the current tracking of containment free volume?

Respcnse: There is no document which contains the current tracking ofcontainmentfree volume. The minimum containmentfree volume is based on Bechtel memo AIB-4742, Containment Pressure Calculationfor ECCS dated 9'30'74, (1,937,733) with the 2% calculation accuracy removed (1,899,738)and rounded down to the nearest thousand (1,899,000).

c) Please provide a discussion / input into how additions / modifications within the containment have been tracked through the years with respect to impact on containment free volume.

Response: In the early 1980 's, additions and deletions of the containment materials andfree volume were handled on an informal basis. In the late 1980's, formal tracking ofmaterials in containment and containmentfree volume was accomplished by procedure EN 21065, Containment Alass Tracking. Revision 1 of this procedure is the current revision.

d) Please provide the revised free volume detennination or evaluation which was performed for the SG Replacement. It was noted in letter SGRP-92-1732 that 48.9 tons of steel were added to containment as a result of the SG Replacement but it is not clear how this was addressed.

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ADMINISTRATIVE CONFERENCE NOTES April 2,1998 Response: 1here is no specific documentationfor the determination or evaluation of changes to the

, containment free volume that occurred during the steam generator replacement project. The change to the free volume was probablyjudged to be insigmficant at that time. Tracking of the materials added to containmentfor H2 generation concerns was accomplishedper EA1F-92-057, Supplement 1, Rev.1.

Follow-up Question:

a) is there any program that tracks the containme J free volume reductions / additions on a cumulative basis?

-Response: No.

2.

Topic: Compliance with NRC Regulatory Guide 1.11 (Kent Russell)

Background:

NRC Regulatory Guide 1.1i states that instrument lines penetrating or connected to i

primary reactor containment should be sized or orificed to assure that in the event of a postulated failure of the piping or of any component in the line outside primary reactor containment during normal reactor operation the leakage is reduced to the maximum extent practical consistent with other safety requirements and the potential off-site exposure will be substantially below the guidelines of10CFR100.

FSAR Section 5.2.8.2.1 states that the containment pressure instrumentation lines are sized or orificed on the inside of the containment such that the response time of the transmitters remains within an acceptable level while in the unlikely event of instrument line or transmitter housing failure, the leakage is reduced to the minimum extent practical.

Questions:

a) is there a calculation or other documentation supporting the RG 1.11 requirement and the FSAR statement relative to the amount of leakage expected from the containment pressure instrumentation lines in the event of a rupture outside containment?

Background continued: NRC Regulatory Guide 1.1I states that instrument lines penetrating or connected to primary reactor containment should be provided with an isolation valve capable of automatic operation or remote operation from the control room or from another appropriate location.

FSAR Section 5.2.8.2.1 takes exception to this portion of the Regulatory Guide for the containment pressure instrumentation lines, stating that a single manual shutoff valve is used in each line. This J

exception is addressed in correspondence between the AEC and the licensee on the following dates:

02/16n3,08/26/74,08/28n4,11/01/74 and 06/16n5. None of this correspondence documented acceptance by the AEC or NRC of the exception taken to RG 1.11.

Questions:

b) Has the licensce's position on RG 1.11 been explicitly acepted by the NRC7 If so, please identify where this is documented.

Response: Deferred to F9'9K PAGE 3

ADMINISTRATIVE CONFERENCE NOTES April 2,1998

3.,

Topic: CST Inventory Requirements (Gordon Chen) j

Background:

The Tech Spec requirement of minimum CST inventory will be changea from 150,000 gallons to 185,000 gallons (PTSCR 2-20-97). Calculation 97 CST-01999-M2 (dated 1 98) was performed to support this change. This calculation identifics the required CST volume to Lnaintain the RCS in Hot Standby for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> or to cooldown to 300 decrees F in the event of Loss of Off-site Power.

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FSAR Section 10.4.5.3 will be changed per resolution of UIR #2870 to indicate that 115,000 gallons i

of water are required to maintain the plant in Ilot Standby for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. It is not specified for Loss of Off-site Power condition.

Technical Specification 3.4.1.2 requires at least one RCP in cach loop to be operable and at least one loop should be in operation in Hot Standby (Mode 3). Upon Loss of oft-site Power, the plant is required to be in Hot Shutdown (Mode 4) if the requirements above can not be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

1 Questions:

a) Does the below statement in the Bases of Tech Spec 3/4.7/1/3 apply only to the Loss of Off-site Power condition?:

"The minimum water volume is sufficient to maintain the RCS at HOT STANDBY condition for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with steam disciarge to atmosphere."

b) What is the basis for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> in Hot Standby following a Loss of Off-site Power?

)

c) is Reactor Coolant Puup operation considered in calculation of the minimum required CST inventory? Why/why not?

d) Based on the requirement of Tech Spec 3.4.1.2 above, the plant should be in Mode 4 if the Operable condition is not expected to be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Is it necessary to reserve the volume required to cooldown the RCS to 300 degrees F during the 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> the plant is in Hot Standby?

Response: Deferred to v9'98 4.

Topic: EBFS System Scismic duct supports (Roger L. Mauchline) 4

Background:

Drawing series 25203-29126 for the EBFS ductwork has a note indicating duct supports, marked with a symbol consisting of a letter "S" within a shield, are "scismic hangers as detailed by Bechtcl" Questions:

a) Arc there Bechtel details or instructions for these hangers?

Response:.\\filistone is lookingfor Bechtelinstructionsfor seismic ducts.

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ADMINISTRATIVE CONFERENCE NOTES April 2,1998 b) Are these scismic hangers also detailed on Beever Shop Standards, Spec. 506 which are also referenced on the same drawings?

Response: There are details of scismic ducts attached to Specification 7604-506. These have been requested by RAI.

S.

Topic: Valves 2-MS-432A and 2-MS-432B (Louis Mscichowski)

Background:

Tier 3 is reviewing an NCR that deals with valves 2-MS-432A and 2-MS-432B.

These valves cannot be found in the PMMS component database nor can they be found on the Main Steam P&lD 25203 26002 Sheets 1 through 5.

Questions:

a) llave these valves been removed from the system?

Response: Yes they have been removedfrom the system. They were drain valves and were removed in 1983.

b) If they have been removed. what is the mechanism that removed them?

Response: The PDCR that removed them will be identified later.

6.

Topic: PDCE MP2-89-014 (Candace Segar)

Background:

PDCE MP2-89-014 replaces melamine torque switches with fibrite torque switches.

Questions:

1 I

a) What document scismically qualified these replacement torque switches?

I Response: Limitorque provided a certificate ofcompliancefor the SAfB series valve actuatorsfor Millstone Unit 2. RAI (1320) issued requesting this COC.

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Topic: Main Feedwater Fystem performance during MSLB cvent (Gordon Chen) i

Background:

The Main Feedwater system flow response following the initiation of a Main Steam l

Line Break ( MSLB ) accident is modeled in the current Analysis of Record: (1) to double feedwater flow to the affected steam generator from the initial value, and (2) to coast down in 5 seconds j

following a Main Steam Isolation signal. These assumptions were applied to sensitivity study for various power lewis.

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ADMINISTRATIVE CONFERENCE NOTES April 2,1998 Questions:

a) What document (Evaluation or Calculation) determines that an increase to double the initial flow is bounding for MSLB events? Note this applies to various power levels with corresponding initial Feed Pump speeds and Flow Control Valve positions.

b) What document provides the basis for FW liow coast-down time to be less than 5 seconds?

c) is there a performance test, startup test or record of plant transient that can verify the FW flow coast-down time l

Response: NNECo 's responseJbr Question (7) was E-mailed to Parsons on,f-2-98.

1 8.

Topic: Review of engineering specifications SP-EE-292 and SP EE-105 (Jim Collins)

Background:

During the tier 3 review of engineering specifications for large motors it was noted that some specifications contain requirements defining the capability of the motor to withstand a bus transfer (i.e., SP-EE-292, 4000V AC induction motor for Low Pressure Safety Injection Service, section 4.3). During the review of SP-EE-105, Spare Reactor Coolant Pump Motor, it was noted that j

no such requirements were contained within the specification.

i Questions:

a) Were requirements contained in the original project specification for the motors presently in service?

i b) If so, what is the specification number and why was it deemed unnecessary to be included in SP-EE-1057 c) If the requirements were not included, what was thejustification and how was it documented?

Response: NNECo responded that the requirement was indeed included in the original Byron Jackson specification IT 3586 paragraph 19 and also in paragraph of SP-EE-105 and the Instruction. Manual PTE] 11523 Revision 1. Final review ofSP-EE-105 confirmed the requirement to be as stated by NNECo.

9.

Topic: PDCR 2-231-76 (Robert Moyct)

Background:

PDCR 2-231-76 installed a maintenance opening in a concrete wall of the Enclosure Bldg.

Questions: Please identify:

a) The structural drawing for the wall.

Response: Structural drawings are 25203-11131 and 11116 sect E.

b) DCN's issued for the mod.

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ADMINISTRATIVE CONFERENCE NOTES April 2,1998 Resposise: No DCN's issuedfor mod, drawing's not updated.

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I c) Amount of reinforcing cut.

Response: Cut rebar unknown, field ualkdown will try to identop and NU will respond 4-7-98.

Continuedto 4/9/98.

l d) The original structural calculation for the wall.

Response: Still searchingfor original calc. Continued to 4.'9'98 c) The calculation / evaluation tha' addresses the structural integrity of the wall with the opening installed.

Response: No calculationfor mod:Jied condition..

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10. Topic: DR-313 (NNECo requested)

Response: NNECo pre nted their revised response to DR-313.

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Topic: Containment gas and particulate radiation monitor sampling noules (John Wilkens)

Background:

I am unable to locate the installation details for the containment gas and particulate radiation monitor sampling noules, identified as RN-8123 and RN-8262 on P&lD 25203-26028 Sht.

1. Drawing 25203 29103 is a vendor generic drawing. Drawing 25203-28402 Sht. R03 does not show noule installation detail. Isometric drawings 25203 28408 Sht. 898 and 906 do not show sample noule installation details.

l Questions: Please identify:

a) Please identify documents u hich show the installation details for RN-8123 and RN-8262 Response: Drawing 25203-29103 Sht 49 shows installation detailfor the referenced sampling no::les.

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Topic: Containment high range radiation monitor location and installation detail (John Wilkens)

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l ADMINISTRATIVE CONFERENCE NOTES April 2,1998

Background:

The containment high range radiation monitors RE-8240 and RE-8241 are mounted

,in containment 14'6 Elev. (FS AR 7.5.6.1.2.1.j) on the outside of the biological shield wall in the vicinity of the electrical penetration area (FSAR 7,5.6.1.3). These detectors are not shown on instrurt:nt location drawing 25203-28014 -INSTRUMENT LOCATION CONTAINMENT PLAN EL.14 6" & 38'6". Also, installation details for the detectors cannot be located.

Que< tions: Please identify:

a) Please identify documents w hich show the location and installation details for detectors RE-8240 -

and RE-8241.

Response: Deferred to 4/9/98.

l 13.

Topic: Re-addressing Topic #1 of 3/31 (NNECo)

Background:

. Millstone Unit 2 has modified their response to question 1.a).of 3/31/98 (Andy O'Connor): "Has the Unit 2 Alternate Shutdown Panel and its isolation transfer switch capability ever been functioaally tested either during the original startup or since that time? If so where would the i

l documentation be found?"

Questions: N/A Response: It was originally reported that the inchvidualfunctions are tested, not the panel in its entirety; however, one channel (the channel cahbrationfor the reactor cold leg ternperature) had not been tested.

The original response indicated that no procedure was in place and a CR had been generated.

It has been discovered that SP-2402 J is used to test reactor cold leg ternperature, in addition to the Pressuri:er Pressure Low range originally reported.

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ADMINISTRATIVE CONFERENCE NOTES April 7,1998 DATE:

4/7/98 PURPOSE:

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

1.

N/A LIST OF A TENDEES:

NNECo NRC NEAC Parsons Due to scheduling conflicts, tinis Conference was canceled.

PAGE 1

ADMINISTRATIVE CONFERENCE NOTES April 9,1998 DATE:

4/9/98 PURPOSE:

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

LIST OF ATTENDEES:

NNECo NRC NEAC Parsons This conference was canceled because oflack of NRC representation.

PAGE 1

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ADMINISTRATIVE CONFERENCE NOTES April 14,1998 DATE:

4/14/98 PURPOSE:

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

1.

Resision of Millstone Unit Nos.1,2, and 3 Physical Security Plan f

2.

DGV Calculation 97DGV-01849M2 and EWR 96-105, HVAC Calculation Initiative 3.

NCR 2-94-246

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4.

NNECo response to RAI-1074 5.

Compliance with NRC Regulatory Guide 1.11 6.

CST Inventory Requirements

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7.

NU's revised response to DR-Oll3 8.

Fire Shutdown Panels C-9 and C-10 9.

Containment Pressure Impulse Lines

10. PDCR 2-231-76

11. PDCR 2-122-81

12. Model Numbers for ilS-5276A & HS-5279A

13. ER-27-93 J

14. SP-M2-ME-0003, Revision 0 15.112 Purge System LIST OF ATTENDEES:

NNECo NRC NEAC Parsons i

Ray Necci John Nakoski Wayne Dobson Roy Terry Don Marks Stu Thickman Kent Russell Rich Ewing Gary Jackson Farid Elsabec Gordon Chen Cris Cristallo Dan Wooddell C:.ris Scully Andy O'Connor Mike Short Ray Thomas Clark Maxson Jim Giova Bob Skwitz Dale Pruitt Mark Suprenant Roger Miller Jim Nicholson William Clemenson Margaret Skorupski Frank Cobb Mike Cavanaugh William Keegan Ken Fox Mike Akins 1.

Topic: Revision of Millstone Nuclear Power Station Unit Nos.1,2, and 3 Physical Security Plan (Jim Giova)

Hackground: ACR 11878 identified an opening of approximately 14" x 32" in the Auxiliary Building floor at Elev. 38'-6" w hich connects the spent fuel pool compartment at Elev. 38'-6" to the Health Physics Lab at Elev.

14'-6" This opening is located between the flanges of cask crane steel building column N.3-17.2. This opening is between a Protected Area and a Vital Area. ACR 11878 stated that this opening does not create a security concern.

2 DCN DM2-S-0378-96 states that the opening is larger that 96 in and therefore requires a security barrier. This DCN also states that an existing screen barrier is in the opening, however Security has determined that this existing screen barrier is not adequate. DCN DM2-S-0378-96 implemented a repair which welded a horizontal

%" steel plate between the flanges of column N.3 17.2 to totally close the opening. The plate is positioned vertically such that it is flush with the underside of the bottom of the concrete floor slab at Elev. 38'-6" AWO M2 96 03567 documents that this repair has been performed.

PAGE 1

ADMINISTRATIVE CONFERENCE NOTES April 14,1998 Removal of the existing screen barrier and replacing it with the %" steel plate impacts the Physical Security Plan.

The,1CAVP is not normally privy to the Physical Security Plan, and cannot readily verify that the Plan has been resised to reflect this change to the barrier.

Question:

a) Can NNECo provide some documentation to substantiate that the current resision of the Physical Security Plan reficcts the change implemented by DCN DM2-S-0378-967 Response: Securitv representatives stated that the Physical Security Plan has layout drawings ofeach Vital itrea which show the physical barriers and intrusion detection devicesfor each area. Security stated that the new steelplate added by the subject DCN meets the definition ofa Physical Barrier as defined by 10 CFR 73.2, Definitions. Finally, Security stated that the Physical Security Plan does not need to be revised because the existing la>vut drawingsfor this area already show a physical barrier at this location, and the steelplate complies with the 10 CFR 73.2's definitionfor a physical barrier.

10 CFR 73, Physical Protection ofPlants and Afaterials, section 73.2, Definitions defines a Physical Barrier as (2) Building walls, celhngs, andfloors constructed ofstone, brick, cinder block, concrete, steel or comparable materials (openings in uhich are secured by grates, doors, or covers ofconstruction andfastening ofsufficient strength such that the integrity ofthe wallis not lessened by any opening), or walls ofsimilar construction, not part ofa building, provided with a barbed topping describedin paragraph (1) ofthis definition ofa height of not less than 8 feet: or (3) <tny other physical obstruction constructed in a manner and ofmaterials suitablefor the purposefor which the obstruction is intended.

2.

Topic: DGV Calculation 97DGV-01849M2 and EWR 96-105, HVAC Calculation Initiative (Bill Clemenson)

Background:

The scope and status of EWR 96-105, HVAC Calculation Initiative has been discussed with NNECo on numerous occasions. A reading of the scope (Disposition) associated with EWR 96-105 is to scope out ventilation calculations w hich need resision and to revise and or replace the applicable calculations. Per telecon on 2/3/98, discussion topic 6, in response to our questions on the scope of this HVAC initiative, NNECo stated that Phase I and II of this initiative is completed and resulted in 16 resised cales for the Diesel Generator ventilation System.

Calculation 97DGV-01849M2 (DGV Duct Sizing and Pressure Drop Cales) was issued 9-10-97 to supersede 2N2018, 2N20-19,2N20-20,2N70-23. As stated in the. Purpose, this new calculation does not address changes that have been made to the DGV flow or configuration but was issued only to gather and store this historical (albeit incorrect) infonnation. In addition, it is stated that system test and balance tests which are performed following changes in the ductwork configuration are the basis ofsystem acceptability and equipment adequacy. And finally it is stated that prior to any future design modifications a new or resised calculation reflecting the current ductwork will be required.

It is not clear how we should proceed with regard to review of these new resised calculations which are simply a repackaging ofold inaccurate cales. Tic original problem of hasing calculations u hich do not reflect current plant configuration still exists. Since Phase I and Phase 11 of this ilVAC initiative is completed, is it safe to assume that there will be no more work performed under EWR 96 105 on system 2315E ventilation calculations?

Questions:

a) Please confinn that Calculation 97DGV-01849M2 is the new calculation of record for Diesel Generator Building Ventilation Dxtwork sizing and pressure drop calculations.

Response: Confinned PAGE 2

ADMINISTRATIVE CONFERENCE NOTES April 14,1998 b) Please provide a listing of TAB docmnents that we can request which provide basis for DGV system acceptability.

Response: EN21063A andEN21063G c) Please confinn if work on sys%n 2315E ventilation calculations as scoped by EWR 96-105 is complete.

Response: Confirmed I

3.

Topic: NCR 2 94-246 (William Keegan) l l

\\

l Hackground: NCR 2-94-246 is being reviewed with respect to the classification changes of LS-8769 & LS-8770 i

as Category 1. The level switches were replaced from a McDonnell & Miller to Magnetrol. A review of work

)

orders indicates LS-8769 was replaced by WO M2-87-08934 and LS-8770 was replaced by WO's M2-87-08933 &

M2-89-01990. MEPL CD-1131 classified the switches as non-Category I and MEPL CD-1494 & CD-1670 reclassified the switches as Category 1. RIE TSE-MP02-95-0068 evaluated the change from McDonnell & Miller to Magnetrol.

Questions:

l a) Were the replacement level switches QA Category 17 t

I l

l Response: The replacement switches were non-QA Category I as required by the current 1988 MEPL in l

effect at the time ofinstallation, j

l b) If the switches were not Category 1, was there a commercial grade dedication package or a purchase order documentation upgrade for the replacement switches 7 i

t I

Response: NNECo responded that all existing, previously purchased level switchesfor this application, were upgraded to Category 1 through the NCR process perprocedure NGP 3.05. A commercialgrade dedication package orpurchase order upgrade was not utili:ed.

l c) What are these documentation package numbers?

Response: NNECo stated that the documentationfor the installed component upgrades was included in the NCR disposition. NNECo hadpreviously supplied the complete disposition with RIE 75E-MP02-95-0068 via RAl-1241. The RIE evaluated replacement equivalency but it is not apparent that the RIE was a processfor upgrading an existing installed component. This upgrade process via the NCR andRIE was not apparent by a review ofNGP 3.05, rev. 7.

PAGE3

ADMINISTRATIVE CONFERENCE NOTES April 14,1998 d) Do the currently installed switches have the required QA Category I documentatior.?

Response: NNECo respondedyes, via the NCR disposition.

r 4.

Topic: NNECo response to RA!.1074 (Joe Groncki)

Background:

NNECo response to RA11074 stated in part, ' Seismic documentation for the remaining 84

)

components as indicated in M2-RAl-01074 were not located in Nuclear Document Services (NDS). CR No. M2-f 98-0868 was generated to track resolution for these items. These :: mall (21/2"and under) manual valves and l

check valves, filter elements, and strainers may have been considered seismically rugged / insensitive items during 1

l procurement in the 70s, therefore seismic documentation may not exist."

Questions:

a) What is the current status of the search for the missing seismic documentation? Has it been determined that this documentation does not exist?

Response: The numberfor the AR associated with CR No. M2-98-0868 is AR 98006613-01. The estimated completion datefor the corrective action plan is 1/30'98.

5.

Topic: Co.mpliance with NRC Regulatory Guide 1.11 (Kent Russell)

Background:

NRC Regulatory Guide 1.11 states that instrument lines penetrating or connected to primary reactor containment should be sized or orificed to assure that in the event of a postulated failure of the piping or of any component in the line outside primary reactor containment during norr i reactor operation the leakage is reduced to the maximum extent practical consistent with other safety reqmrements and the potential off-site exposure will be substantially below the guidelines of 10CFR100, FSAR Section 5.2.8.2.1 states that the containment pressure instrumentation lines are sized or orificed on the inside of the containment such that the response time of the transmitters remains within an acceptable level while in the unlikely event of instrument line or transmitter housing failure, the leakage is reduced to the minimum extent practical.

Questions:

a) Is there a calculation or other documentation supporting the RG 1.11 requirement and the FSAR statement relative to the amount of leakage expected from the containment pressure instmmentation lines in the event of a rupture outside containment?

l I

l Response: There is no calculation or other documented basisfor the cjuantity ofleakage expectedfrom the containment pressure instrumentation hnes.

Background continued: NRC Regulatory Guide 1.11 states that instrument lines penetrating or connected to primary reactor containment should be provided with an isolation valve capable of automatic operation or remote operation from the nntrol room or from another appropriate location. FSAR Section 5.2.8.2.1 takes exception to this portion of the Regum 'ry Guide for the contaimnent pressure instrumentation lines, stating that a single PAGE 4

ADMINISTRATIVE CONFERENCE NOTES April 14,1998 manual shutofT valve is used in each line. This exception is addressed in correspondence between the AEC and the licensee on the following dates: 02/16/73,08/26/74,08/28/74,11/01/74 and 06/16/75. None of this correspondence doeurrented acceptance by the AEC or NRC of the exception taken to RG 1.11.

Questions:

b) lias the licensee's position on RG 1.11 been explicitly accepted by the NRC7 If so, please identify where this is documented.

Response: l'es. The NRC accepted the position on RG 1.11 in a letter dated 8/1/75.

l 6.

Topic: CST Inventory Requirements (Gordon Chen)

Background:

The Tech Spec requirement of minimum CST inventory will be cha'iged from 150,000 gallons to 185,000 gallons (PTSCR 2-20-97). Calculation 97 CST-01999-M2 (dated 1-21-98) was performed to support this change. This calculation identifies the required CST volume to maintain the RCS in Hot Standby for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> or to cooldown to 300 decrees F in the event of Loss of Off-site Power.

l FSAR Section 10.4.5.3 will be changed per resolution of UIR #2870 to indicate that i15,000 gallons of water are required to maintain the plant in llot Standby for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. It is not specified for Loss of Off-site Power condition.

Technical Specification 3.4.1.2 requires at least one RCP in each loop to be operable and at least one loop should be in operation in llot Standby (Mode 3). Upon Loss of Ofr-site Power, the plant is required to te in flot Shutdown (Mode 4) if the requirements above can not be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Questions:

a) Docs the below statement in the Bases of Tech Spec 3/4.7/1/3 apply only to the Loss of Off-site Power condition?:

J "The minimum water volume is sufficient to maintain the RCS at ilOT STANDBY condition for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with steam discharge to atmosphere."

b) What is the basis for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> in llot Standby following a Loss of Off-site Power?

c) is Reactor Coolant Pump operation considered in calculation of the minimum required CST invemay?

Why/why not?

d) Based on the requirement of Tech Spec 3.4.1.2 above, the plant should be in Mode 4 if the Operable condition is not expected to be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Is it necessary to reserve the volume required to cooldown the RCS to 300 degrees F during the 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> the plant is in liot Standby?

Response

Part (a), (b), and (c): The design capacity for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of Hot Standby is for the Decay Heat removal, not including the pump heatfrom the Coolant Pumps. It is based on the CE design criteria N-SE-66, dated 6'69.

This has been reviewed and concluded to be acceptable by NRC through SER, dated 3/1074 Part (d): &. The CST volume requirement is not based on the Hot Standby operation followed by the cooldown to 300 degrees F. This requirement ofcapacity is specyied in hRC Branch Technical Position RSB 5-1, dated 7/1981. It is not applicable to Millstone 2, which was desigt.ed before RSB 5-1 was published.

PAGE5

ADMINISTRATIVE CONFERENCE NOTES April 14,1998 7.

Topic: NU's revised response to DR-0113 (Dan Wooddell)

Background:

NU's revised response to DR-0113 does not contain all test documentation expected by Parsons.

In addition, NU did not provide a revised response to the PDCR 2-52-95 safety evaluation issue, as discussed during the 1-19-98 meeting to discuss high level DRs.

Questions:

a) N/A Response: NU was informed of the documentation Parsons requires in the revised response to this DR. NU does not intend to provide a revised response to the PDCR 2-52-95 safety evaluation issue. NU stated that a new safety evaluation is being performedfor all Containment Sump isolation valve pressure locking concerns.

8.

Topic: Fire Shutdown Panels C-9 and C-10 (Andy O'Connor/ Dan Wooddell)

Background:

N/A Questions:

a) llave the Unit 2 Appendix R Fire Shutdown Panels C-9 and C-10 and their associated ' Bottle-Up Panels" (C70A and C708) ever been functionally tested either during the original startup or since that time?

b) If so, where would the documentation be found?

Response: It was reported that the testing was done during the Appendix R Backfit in Service Test. TheIn Service Test, isfound in Station Test Procedure T-86-36 Rev. O. An RAI requesting this information has been submitted on this date.

9.

Topic: Containment Pressure Impulse Lines (Dan Wooddell)

Background:

N/A Questions:

a) What method is used to provide reasonable assurance that the containment pressure impulse lines (atmospheric) are not blocked or restricted?

Response: h Uperforms a periodic calibration ofthe transmitter loops. This surveillance verifies that the lines

- are not blockedfrom the containment isolation valve to the transmitter. In addition, the response of the transmitters is monitored during the periodic IlltT. NU stated that they have not identified a recogni:able methodfor blocking the sensing lines with the exception ofcrushing the hnes during construction. Crushing a hne during construction would be detected during the post maintenance inspection.

b) What procedure is used to control this evolution?

i Response: NU does not have a single procedure to veri [v the pressure impulse lines are not blocked or restricted.

PAGE6

ADMINISTRATIVE CONFERENCE NOTES April 14,1998 10.

Topic: PDCR 2-231-76 (Robert Moyer)(Continuation of Topic #9 from 4/2/98)

Hackground: PDCR 2-231-76 installed a maintenance opening in a concrete wall of the Enclosure Bldg.

Questions: Please identify:

a) [ Answered by NNECo on 4/2/98]

b) [ Answered by NNECo on 4/2/98) c) Amount of reinforcing ent.

Response: NNEco walkedslown the maintenance opening in the concrete wall to deternnine the number ofcut rebar. The cutout in the wallis $" deep,1& wide, and 20"high. There is one cut horizontal rebar approx.10" from the bottom of the opening. There is one vertical cut rebar on the extreme right ofthe opening that could be seen visually. No other vertical cut rebar could be seen, but it is a congested area and acctrdung to the drawings there could be one additional cut rebar since the rebar should be on 18" centers.

d) The original structural calculation for the wall.

Response: No specific calculation wasfound The wallis inc/udedin the structural calculatwn modelfor the Enclosure Bldg., but no credit is takenfor load carrying capabihty ofthe wall.

i l

l c) [ Answered by NNECo on 4/2/981 11.

Topic: PDCR 2-122-81 (Roger Miller)

Background:

This PDCR replaced a transformer (UL63) and a molded case breake 32-B6166, feeding a heat j

trace panel. The breaker was ir: stalled in Safety Related MCC B61, According to the PDCR, the breaker was l

purchased commercial grade.

\\

Questions: Please identify:

a) Was the breaker that was installed actually purchased as commercial grade?

Response: Yes, purchased commercial Grade.

b) Was there a program in place to upgrade or qualify commercial grade breakers for use in safety related panels at the time that this breaker was installed?

Response: Yes, but this breaker was not dedicated under that program.

c) Was there an evaluation performed to allow installation of a commercial grade breaker in this instance?

PAGE 7

ADMINISTRATIVE CONFERENCE NOTES April 14,1998 Response: 'No,the breaker was installed without proper evaluation.

d) Would you provide copies of the purchase requisition and maintenance work orders that purchased and installed this breaker?

Response: Breaker was drawnfrom stock. No purchase repetition available. Breaker installed as non-safety, Alaintenance ll'ork Order - AIR 273881 (will request by RAI). The fact that the breaker was not properly dedicated is addressed by UIR 2787, a Corrective Action CR Al2 97 0881 was issued on 3/28/97 and the breaker was dedicated as installed by Commercial Grade Dedication package 207 737. All of which will be requested by ll41.

12.

Topic: Model Numbers for HS-5276A & liS-5279A (Dale Pruitt)

References:

1.

PDCR MP2-092-76, Auxiliary Feedwater Valve Handswitch PDCR MP2-88-048 upgraded HS 5276A & HS 5279A from model CR 2940 to SB 1 (Spring return type). Drawing 25203-32012, Sheet B, Revision 6, shows these switches as Models CR 2940.

A search of the PMMS Database did not locate a document that changed these switches back to model CR 2940. Also the Logic drawings,25203-28105, Sheets 22A and 22B do not state spring retum types as do other logic drawings (e.g. 25203-28105 Sheets 21 & 31).

Questions:

n) Was this intended to be a temporary modification?

Response: No, the modification was never installed it is understood the documentation of this particular modification may be confusing. It will be revi:wed and consideration ofissuing a CR will be made.

b) If not what document changed the switches back to CR 2940 models.

Response: The modification was never installed thus there was no switch back to the CR 2940 model. Current documentation depicts the model CR 2940 is installedfor each ofthese switches.

c) If these were not changed back to CR 2940 what is installed?

Response: Perfield walkdown model CR2940 are installed PAGE 8

ADMINISTRATIVE CONFERENCE NOTES April 14,1998 13.

. Topic:.ER-27 93 (Ray Thomas)

Background:

The Basis Document for Surveillance Procedure SP 2402B lists SER-27-93 under " Basis Information" on page 8, the following sentence concerning calibration of safety-related transmitters. "Per SER-27-93 transmitters are no longer cycled over their calibrated range three times prior to taking As Found data."

Questions:

a) is SER-27-93 an INPO document?

i Response: SER-27 93 is an INPO document.

i b) What was Millstone's response to SER-27 937 Response: Afillstone 's response to SER-27-93 was based on Unit 1 's (BWR) response in i993 based on a review and evaluation ofthe irssons Learnedfrom this event. The Lessons Learned are: " Note: The technical aspects of this event are primarily ofinterest to personnel operating boiling water reactors. However, aspects regarding proper collection and use oftest datafor monitoring equipment performance and thorough use ofindustry operating experience are applicable to allpourplantpersonnel." Afillstone Unit I addressed this issue with a PIR (equivalent to a CR) on procedure SP 105L Afillstone Unit 2 Revised Plant Procedures including SP 2402B to be in compliance with SER 27-93.

I c) Over what time interval did Surveillance Procedure SP 2402B's calibration use cycling over the transmitter's calibrated range three times prior to taking As Found data exist?

Response: Surveillance Procedure SP 2402B 's calibration used cycling over the transmitter's calibrated range three times prior to taking As Found data before 1993. It was noted that in order to cahbrate the 1500 to 2500 Psig Pressuri:er Pressure transmitters calibrated via SP 2102B the Foxboro transmitter would be removedfrom the process and then calibrated. 7his type ofcalibration was considered by Afillstone to be one more cycling of the transmitter in addition to the three later cycles mentioned in SER 27-93. (" Equipment should not be exercisedprior to testing when the results are usedfor evaluatmg equipment operability orfor trending performance. ")

t l

d) Was an evaluation performed to see if the results obtained from calibrations performed using a cycling over I

the transmitters' calibrated range three times prior to taking As Found data affected any Technical l

Specification Limits?

l PAGE 9

l ADMINISTRATIVE CONFERENCE NOTES April 14,1998 Response: No evaluation was performed to see if the results obtainedfrom cahbrations performed using a cycling over'the transmitters calibrated range three times prior to taking As Found data affected any Technical i

I i

l Specification Limits because it was believed that the SER 27 93 mainly applied to BilR Plants and that enough 1

margin exists in the "As Found" data due to the Accuracy Term, Hysteresis, Deadband, Temperature Efects and i

)

other eJJects to assure Millstone that the values obtainedprior to 1993 were still conservative and valid It was also noted that the practice ofcycling over the transmitters cahbrated range three times prior to taking As Fowtd l

data was a standard industry practice.

c) What steps were taken to prevent the use of these results in any future calculation or " Drift Study"7 Response: No steps were taken to prevent the use of these results in anyfuture calculation or " Drip Study" because as stated above the pre I993 Drip data wasfelt to be conservative. ivofurther steps are planned Parson's Response to Millstone (by Ray Thomas): The main reasoning of this question was to assure that inaccurate data obtainedprior to 1993, when the transmitters were cycled over their entire range 3 times before recording an "As Found" value, was not prohferated into today's decision making regarding accuracy. Cycling of the transmitters up and down three times prior to recording any Dnp data skews the results andlowsrs the "As Found" value. An Erample ofwhere this data has been used, is in the "Drt) Evaluations"for a 30 month Dnft interval 14.

Topic: SP-M2-ME-0003, Revision 0 (Gary Jackson) i

Background:

Project Specific Design Specification for Pim Rupture Analysis Criteria Outside the Reactor Building SP-M2-ME-0003, Revision 0, Section 6.1.3.1 %gh Energy Piping Systems" of this document states "High Energy systems may also be classified as moderate energy if the total time of system operation is less than 1% of the total plant operating time".

Questions:

a) What is the Design and Licensing Basis for this statement?

b) What is tiu source of this statement Response: Original Design Basis as contained in Amendment 17 states that those lines oflow usage do not require break postulation. The 1% rule is part of a probabilistic approach which is based on those lines where their usage is such that the probability ofa break occurring in these lines is greater than 1 X 10E-6 15.

Topic: H2 Purge System (Gary Jackson)

Background:

DCN No. DM2-00-0360-97 and DM2-00-0485-97 perform modifications to the H2 Purge System.

l Questions:

a) What PDCR (s) are associated with these DCN's?

1 PAGE 10

v o

ADMINISTRATIVE CONFERENCE NOTES April 14,1998 Response: The DCR !s All-970%

Background Continr.ed: Page i of both DCN's - Block 11, the words " Temp Mod." Are inserted in this block and reference M2-97008.

b) The justificatit,n section for this modification reads as though this is a permanent modification. However, Block 1.1 of Page i indicates this as a temporary modification. Is this a temporary modification and if so what will be considered the permanent modification?

Response: This is a permanent mod:Jication.

c) What is M2 9700087 Respcase: 7he DCR is A/2-9700s I

PAGF 11

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