ML20216B068
| ML20216B068 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/07/1998 |
| From: | Curry D AFFILIATION NOT ASSIGNED |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUM2-PPNR-1333, NUDOCS 9804130360 | |
| Download: ML20216B068 (24) | |
Text
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PARSONS Daniel L. Curry, Vice Ptosnient. Nuclear Serymes Parsons EnerDy & Chemicals Group Inc.
2675 Morgantown Road
- Reading, Pennsyivania 19607 * (610) 855-2366 + Fax. (610) 855-2602 April 7,1998 Docket No. 50-336 Parsons NUM2-PPNR-1333-L U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 Millstone Nuclear Power Station Unit No. 2 Lndependent Corrective Action Verification Pronram (ICA_YP)
Gentlemen:
This letter traramits summaries of telephone conferences between Parsons Power Group Inc., the U. S.
Nuclear Regulatory Commission, NNECo and NEAC on March 17, March 19, March 24, March 26, March 31,1998.
Please call me at (610) 855-2366 if you have any questions.
Sincerely,
. hbAdlV
- g. Daniel L. Curry 1
Parsons ICAVP Project Directer DLC:djv Atuichments 1.
Telephone Conference Notes from March 17,1998 2.
Telephone Conference Notes from March 19,1998
\\\\
3.
Telephone Conference Notes from March.24,1998 4.
Telephone Conference Notes from March 26,1998 g
5.
Telephone Conference Notes from March 31,1998 QO cc:
E. Imbro (2)- USNRC J. Fougere - NNECo H. Eichenholz - USNRC Rep. Terry Concannon - NEAC R. Laudenat - NNECo Project Files 9004130360 980407 PDR ADOCK 0500 6,
P PPNR1143 doc
i ADMINISTRATIVE CONFERENCE NOTES Mrrch 17,1998 DATE:
3/17/98
)
- PURPOSE:
Administrative telephone conference with NNECo, NRC, NEAC and Parsons to discuss:
- 1. Calculation 97-122, Rev. C,1/5/98 2.
Fuel Cask Drop Accident 3.
Steam Generator Tube Rupture Accident 4.
Follow-up to 3/3/98 Meeting 5.
Request to discuss DR-113 [Not on original agendaj l
LIST OF ATTENDEES:
NNECo NRC NEAC Parsons Joe Fougere Ralph Architzel Mike Akins Jim Wheeler Don Marks i
Fred Mattioli John Ihlbish Don Miller (New Britain)
Rich Glaviano Greg Tardif Wayne Choromanski Dave Bajumpa Gordon Chen l
Mike O'Meara Richard Boyd Ray 'Iliomas i
1.
Topic: Calculation 97-122, Rev, C,1/5/98 (Richard Boyd)
Background:
Calculation 97-122, Rev. C,1/5/98; Millstone 2 ECCS System Analysis evaluated l
HPSI and LPSI flows with a variety of scenarios. These flows are then used in the various analysis for response to Chapter 14 accidents.
Under Conclusions. it is stated that ' All recirculation cases were evaluated with the conservative l
assumption that the containment sump screen was partially blocked providing a maximum amount of resistance at the pump suction.'
l Section 4.7 Containmen. Sumn in the calculation does not identify the sump screen in.he model. It indicates the sump is modeled as a pressure suction node at an elevation of -18.9 feet.
Questions:
a) Where is the sump screen modeled?
Response: Calculation 97-071 is the calcu'ation uhich modeled the HPSI and LPS1 systems.
Calculation 97-122 is the analysis based on the model. Section 3.9 of the model calculation models the sump screen as a resistance in the. ne. In response to u hat is the revision level ofthe model, this will be provided 3/19'98 l
Background continued: Under Conclusions, it states that 'the case 2b2a (recirculation with dicscl L
failure, maximum flow) is not credible since it requires two failures (dicsci failure and failure of a LPSI pump to trip on SRAS).'
Section 4.4 Analysis Cases, for case 2B2a does not indicate LPSI operating. Nor is there any recirculation identified with the failure of a LPSI pump to trip on SRAS; this will affect the case of two pumps operating on recirculation considering NPSH requirements.
Questions:
a) Where is the case of a LPSI pump failing to trip on recirculation analped with HPSI pumps running?
PAGE 1
ADMINISTRATIVE CONFERENCE NOTES Mereh 17,1998 Respoase: The sump screen resistance is included in all analysis because it is incorporated in the model as afixed resistance. The statement that the case is not ' credible
- is in regards to the result indicating NPSilratio being less than 0.99; not the case 2B2a.
Background continued: FSAR Section 6.3.2.1 p 6.3 2 states that NPS11 during recirculation is based on ' total flow on one suction header of 2980 gpm, consisting of 1700 gpm for one containment spray pump and 640 gpm for each high pressure safety injection pump. It was assumed that two, high pressure safety injection pumps oparated from one suction header.'
Questions:
a) Is the case of two HPSI pumps and 1 CS pump on a suction header to be analyzed?
Response: The case of2 flPSI and I CSpump operating on one suction header will not be modeled 7here is no 1%1RCR associated with this. The ES4R will be changed based on the results ofall analysis at a later date b) As a follow on question, NECO was asked whether or not the motor horsepower was evaluated, considering the flows in the analysis are higher than previously analyzed?
Response: The response was that it was not evaluated in the calculation, however, previously the motor horsepower was looked at with 690 gym and wasfound satisfactory. This was questioned because theflows in the anaysis are higher than what was evaluated in that case. It was decided to l
respond to this question in the 3/19'98m telecon.
2.
Topic: Fuel Cask Drop Accident (Gordon Chen) [ Holdover from 3/12/98]
Background:
Technical Specification Amendment #172 following the removal of the blocking desices to reclaim 234 fuel storage locations changed the requirement of decay time from 120 days to one year for the spent fuel to be stored within L distance from the center of the cask set-down area whenever a cask is on the refueling floor. According to the letter B14470 that requested the changes.
the cask drop accident has been reanalyzed for the radiological consequences. This was due to the changes in fuel storage capacity in the " targeted footprint arca". It indicates the there are 782 fuel assemblics in the cask drop footprint area.
Questions:
a) What is the calculation number of the reanalysis for the radiological consequences following a fuel cask drop accident to support Tech. Spec. Amendment #1727 Response: The Cale numberis.LT-XLT-1131bt, dated 3/2/93.
b) What is the document that prmides the basis of determining the " targeted footprint area" of 782 fuct assemblies?
Response: The document is 82-232-881bt, Rev. O.
PAGE 2
ADMINISTRATIVE CONFERENCE NOTES M:rch 17,1998 c) What is the value of"L" distance and the basis of this value?
Response: 7he value of"L"is 21 feet. See document ofitem (b)for its basis.
3.
Topic: Steam Generator Tube Rupture Accident (Juan Cajigas) [ Holdover from 3/12/98)
Background:
The following radiological dose calculations were provided for the Tier 2 review of the Steam Generator Tube Rupture Accident (SGTRA):
Calculation xx-xxx-61RA, Revision 0," Radiological Consequences Of A Steam Generator Tube Rupture At Millstone 11", Approved 3/9/83.
Calculation xx-xxx-61RA, Revision 1, " Radiological Consequences Of A Steam Generator Tut c Rupture At Millstone II, Approved 4/29/83.
Calculation xx-xxx 61RA, Resision 2," Radiological Consequences Of A Steam Generator Tube Rupture At Millstonc !!", Approved 5/20/84.
Unnumbered radiological dose calculation entitled "MP-2 Radiological Consequences Of A Steam Generator Tube Rupture At Millstone II", Approved 8/31/94 was also provided. The dose
{
results currently contained in the MP-2 FS AR are obtained from this analysis.
In order to identify system design inputs used in the SGTRA dose analysis, all four calculations were reviewed. Results of this indicate that:
To evaluate the radiological consequences of this event, two separate analysis are performed.
A hydraulic systems analysis is first performed by the NUSCo Safety Analysis Branch using the RETRAN Computer Code to evaluate mass releases for this event. This analysis cvaluates system hydraulic response to obtain mass flow rates and a timing sequence of events for the accident. Results of this analysis are then input into the SGTRA computer code to evaluate activity released to the environment and the resulting radiological dose consequences. Radiological dose calculations for the SGTRA are listed above.
The SGTRA radiological dose calculations only identify radiological assumptions (e.g. fluid activity concentrations, partition factors) and flow rates from the RETRAN analyses. Design inputs such as trip setpoint pressures and valve actuation setpoints (open and re-seat pressures) are not identified in the radiclogical calenhtions. These parameu are listed in FSAR 14.6.3. In order to identify all of & SUTRA system design inputs, the Steam Generator Tube Rupture Accident RETRAN analyses must also be reviewed.
The radiological dose calculations do not always identify the system response scenarios for the radiological cases evaluated. Revision 2 of xx-xxx-61RA cvaluates three new l
radiological cases referencing the RETRAN analysis, but does not clearly explain the basis l
for these new cases. The system response scenario from one of these cases is the one documented in the FSAR. In order to determine the basis for the radiological case results presented in the FSAR, the RETRAN analyses must also be resiewed in order to identify the system response scenarios.
The SGTRA dose results reported in FSAR Section 14.6.3 are evahiated in an un-numbered radiological dose calculation (See calcu'.ation number 4 identified above). It is not clear if this calculation is a design basis MP-2 analysis since it is un-numbered. This calculation uses base dose inputs from calculation xx-xxx-61RA, Resision 0 (1983) which was superseded by calculation xx-xxx-61RA, Revision I, and process data from a 1992 RETRAN analysis which appears inconsistent with the Rev. O SGTRA calculation. It also uses input from calculation xx-xxx-61RA, Revision 2 and references " updated" doses from a 1992 memo w hich was issued years after the previous calculation (xx-xxx-61RA, Rev.2,1984).
I PAGE 3
T i
i[
ADMINISTRATIVE CONFERENCE NOTES Mirch 17,1998 Questions: Please clarify the following:
a) Which radiological calculation (s) provide the basis for the analysis of record?
)
Response: 7he "un-numbered" calculation.
b) In the event that the un-numbered calculation above (4) provides this basis, why does it references calculations that have been superseded? In addition, why is process data (steam flows, break flows) from W2-517-1015-RE Rev.! (1992) used in combination with calculated dose data from xx-xxx-61RA Rev.0 (1983) which appears to use different process data?
Response: Response to be provided at a later conference.
t c) Which RETRAN calculations / analyses are part of the SGTRA analysis of record?
Respcnse: lV2-517-1015-RE Rev.2 1.
Topic: Follow-up to 3/3/98 Meeting -(NNECo) [ Holdover Topic from 3/10/98.]
NNECo to prcvide Parsons and S&L with information on the follewing topics; a) Minor Modifications Safety Evaluation Screening e
Status of S&L DR - 6 b) Setpoint Control c) PI 20 deferment criteria d) Millstone approach on correcting labeling discrepancies e) S&L DR 1007 Response: Deferred until NNECo is prepared to discuss the Topic. NNECo uill announce the discussion date and time by telephone to allparties.
5.
Topic: Request to discuss DR-113 Meeting -(NNECo)
Background:
NNECo requested that a discussion of DR-113 be put on the agenda for the next conference call.
Questions:
a) N/A Response: Parsons will include the discussion of DR-113 on the 3/19'98 Conference agenda.
PAGE 4 j
ADMINISTRATIVE CONFERENCE NOTES Mrrch 19,1998 DATE:
3/19/98 3
PURPOSE:
Administrative telephone conference with NNECo, NRC, NEAC and Parsons to discuss:
1.
Scismic Interaction Program
- 2. FSA4 Chapter 14 Accident Analyscs 3.
Discussion of DR-113 4.
PDCR 2-047-92 Rev 2 5.
Section XI Index 6.
Steam Generator Tube Ruptare Analysis 7.
SP-EE-261, Rev 2,7/1/89 8.
Calculation 97-122 Rev. C,1/5/98 LIST OF ATTENDEES:
NNECo NRC NEAC Parsons Ray Necci John Nakoski Wayne Dobson Fred Mattioli Don Marks Laird Bmster Richard Hoyd
)
Mike O'Meara Fletch Downev
{
Harvey Becman Candace Segar Ken Fox Ken Gabel Greg Tardir ImTy Collier Nabil Juraydini Juan Cajigas i
Rich Ewing Dan Wooddell Dave Bajumpa Roger IIall Mike Smaga Paul Schmitzer Dan Cardinale Rich Glaviano Mike Akins 1.
Topic: Seismic Interaction Program (Roger Mauchline)
Background:
Document MS-35 Revision 2," Criteria and Action To Be Taken For Protection of Safety-Related Components From Damage Non-seismic Components Which Might Occur During A Scismic Event," has criteria that can be used to identify seismic interaction concerns, and also outlines the kinds of measures that can be taken when concern are identified. Seismic interaction is also addressed by the General Implcracntation Procedure, (GIP) w hich was used in the Scismic Evaluation Report. The GIP addresses seismic interaction in two ways: (1) Interactions with safety equipment (GIP item 4.5) and (2) Interactions with safety related raceways (GIP item 8.2.5).
Questions:
a) Was document MS-35 used in plant design and modifications? If so where is the documentation j
of the program or effort for this activity?
Response: Ms35 was usedfor originalplant design engineering evaluations.
b) What criteria and programs were used to address scismic interaction in addition to the GIP inspections implemented in the Seismic Evaluation Report?
l Response: NEO 5.19, which provided additionalguidance on uhat to lookfor (issuedin 19h6), was j'
implementedin conjunction with the design review ofplant changes via PDCR 's.
PAGE 1
ADMINISTRATIVE CONFERENCE NOTES MCrch 19,1998 c) What criteria and programs are currently being used to address scismic interaction for plant modifications?
Response: In conjunction with development ofthe DCMprogram, NGP 5.19 was generated to define the processforperforming Seismic Quahfication Reviews.
Continaation Questions:
Second part of Question A.
What criteria / programs were used between the original plant design and the 1986 time frame?
What seismic categc ry 'i/l criteria was implemented to perform UIR 1769 required 11/1 e
reviews?
What is the status ol.ne UIR 2567 recommendation to evahiate the need to develop a e
program to establish seismic 11/1 adequacy?
Response: NNECo willrespond to these questions on 3'24/98.
2.
Topic: FSAR Chapter 14 Accident Analyses (R. Glaviano)
Background:
Several FSAR Chapter 14 accidents are being reanalyzed by NNECo. Design inputs and completed analyses are to be provided by NNECo to Parsons to enable completion of the Tier-2 review. The current information needs are identified in Attachment B of the bi-weekly progress report dated March 9,1998.
Question:
i a) Please provide an update to the forecast completion dates for the event re-analyses listed in l
Attachment B to the bi-weekly report.
Response: 7he revised 151R Chapter 14 event re-analyses areforecast to be completed by 4'30.
l l
b) Please identify w hich analyses have "mmimal" design input changes and briefly discuss the j
nature of the " minimal" changes.
Response: The remaining design inputs will be provided to Parsons by April 15. A summary ofthe scope ofdesign input changes wasprovided to Parsons.
3.
Topic: Discussion of DR-113. (NNECo)
Background:
N/A Questions:
a) N/A Respense: Def.rred to 3/24198.
PAGE 2 m
ADMINISTRATIVE CONFERENCE NOTES M:rch 19,1998 4.
Topic: PDCR 2-047-92 Rev 2 (Candace Scgar)
Background:
PDCR 2-047-92 Rev 2 re..rences, (in Section 5.1 under Design inputs and Detailed Design item # 3) Main Line Engineering Report No. 89029-003 dated 2/28/90
" Technical Evaluation and Material Assessment for VOTES Hardware Installed on Motor Operated Valves in Nuclear Power Plants." The 2/28/90 date for this report corresponds to Rev 0.
Engineering Report No. 89029-003 was revised with Revision 3 being issued on 9/15/92. PDCR 2-047-92 Rev 2 shows a completed by date of 8/29/95 and approved by date of 9/11/95.
Questions:
a) Were the additional revisions for Engineering Report No. 89029-003 considered during Rev 2. of the modification? If yes, please identify the documentation which shows that Engineering Report No. 89029-003 Revs 1-3 were considered in PDCR 2-047-92 Rev 2, Response: No written documentation has yet been found showing additional revision were considered. NNECo requested the Topic be continued on 3/2488.
5.
Topic: Section XI Index. (Larry Collier)
Background:
ASME Code Section XI, IWA-4700 requires that IWA4000 be used to complete all records for repairs. IWA4340(a) requires that an index be maintained for the record file. IWA-6'40(d) and IWA4340(c) require that repair records, replacement records and reports be maintained for those activities specified in IWA-4000 and IWA-7000. We understand from previous questions and pal responses, that Millstone Unit 2 does not have a complete index.
Questions:
How does Millstone currently maintain and control the required index of repairs or replacements o
of ASME Section XI components? Is this described in a procedure or program? If so, please provide the procedure or program name and number.
Response: There is no programmatic indexfor the plant. There are various references in various procedures that do not lead to a programmatic index.
6.
Topic: Steam Generator Tube Rupture Analysis (Juan Cajigas)
Background:
We understand that the RETRAN an:1 process data from W2-517-1015-RE Rev.) is used in combination with dose data from xx-xxx41RA Rev.0 (note that Rev. O has been superseded) in the *un-numbered" calculation of record because this RETRAN and process data is the same as that used in the xx-xxx41RA Rev.0 analysis. However, inspection of the RETRAN and process data presented in xx-xxx41RA Rev.0 reveals that this data does not match the W2-517-1015-RE Rev.)
data used in the "un-numbered" calculation.
Questions:
a) Please explain why the data can be used in combination with xx-xxx41RA Rev.0 dose data.
PAGE 3
I ADMINISTRATIVE CONFERENCE NOTES Mrrch 19,1998 Response: The "Rev. 0" dose data was obtainedfrom xx-xxx-61RA Rev.2. The "Rev.0" RETRAN data was obtainedfrom It'2-517-403-RE Rev.0. The updated RETRAN data was obtainedfrom li'2-517-1015-RE Rev.2. The analysis is being re-done by Sil'EC.
7.
Topic: SP-EE-261, Rev 2,7/1/89 (Dan Cardinale)
Background:
RAI 0099 requested a copy of all generic non-system unique specifications in seven specified areas. One of the documents transmitted by NNECo was specification SP-EE-261, Rev 2, 7/1/89 " Design Standard for the Modification of Control Pancis "
We recently attempted to determine whether any change documents had been issued against this specification since its issue date, but were unable to locate the base document within the GRITS data base. Our scarch criteria specified all Millstone 2 specifications with installation number 25203.
Questions:
(
a) Does SP-EE 2612 remain the specification for controlling design activities associated with the Main Control Room Panel modifications or has it been replaced by another document? If still i
active, what DCN's have been issued agai:st i:9 If not, what document has taken its place?
l Response: Design Specification SP-EE-261 remains the specification controlling design activity 1
for Control Panels. No DCN's have been issued against this specificationfor Millstone Unit 2.
l t
1 b) is specification SP-EE-261 included in the Millstone data base (GRITS) of design documents? If yes, please explain how it is accessed.
Response: SP-FE-261-2 can be accessedfrom GR17S by using control number 99999 in place of 25203 c) Would Millstone administrative procedures have required conformance of this design specification for any modifications (including labeling changes) to the main control room control panels? Are any such changes waiting to be incorporated at this time? If so, p! case identify them.
Response: Millstone Administrative procedures do not require conformance ofthe specification for changes to the Main Control Panels which may not be covered by the Design Specification. The DCM requires that such changes are subject to 1&C review per NGP 5.25. NGP 5.25 requires a Control Panel Design Review (CPDR) form which is thenfiled with the change document (e.g.) a PDCR. K Fox asked ofwe wanted them to identify change mechanism documentsfor specific labeling changes. It'e advised that we will request specific PDCR numbers at a later date.
PAGE 4
3...
ADMINISTRATIVE CONFERENCE NOTES Merch 19,1998 i
l I
Cdntinuation Questions:
Are labeling changes to Control Panels controlled by PDCR or by some other process such as a Work Order?
Response: Labeling changes can be implemented by a PDCR or by am OPSprocedure. The Ops procedure will be identifiedlater.
8.
Topic: Calculation 97-122 Rev. C,1/5/98 (Richard Boyd)
Background:
Calculation 97-122 Rev. C,1/5/98; ' Millstone 2 ECCS Systen Analysis' evaluated i
HPSI and LPSI flows with a variety of scenarios. These flows are then used in the various analysis for response to Chapter 14 accidents. The motor horsepower requirements for this pump were not identified in this calculation. it appears that the motor will be seriously challenged in several scenarios when the 400 hp motor is expected to provide greater than 455 hp for several of these flow conditions.
Questions:
a) Were the motor horsepower requirements reviewed to ascertain that the pump is capable of satisfying the performance evaluated in this analysis? What documents this resiew?
l Response: The motor horsepower for the llPSI pumps are within the allowable of 460 hp
\\
considering a 400 hp motor with a 1.15 servicefactor.
l 1
PAGE5
ADMINISTRATIVE CONFERENCE NOTES M:rch 24,1998 DATE:
3/24/98
. PURPOSE:
Administrative telephone conference with NNECo, NRC, NEAC and Parsons to discuss:
1.
Seismic Qualification Documentation 2.
Root Cause Evaluations RP 6 3.
RP 4, Corrective Action Program 4.
Containment / Enclosure Purge Supply Ductwork 5.
Discussion of DR-113 6.
Steam Generator Tube Rupture Analysis 7.
PDCR 2-047-92 Rev 2 l
8.
Specification Revisions 9.
NCR's 2-46-78 and 2-51-78
- 10. Seismic Interaction Program j
LIST OF ATTENDEES:
NNECo NRC NEAC Parsons j
Joe Fougere John Nakoski Wayne Dobson j
Fred Mattioli Don Marks 1
Phil Higgin.c Lany Wigley Greg Tardif Richard Boyd George Pitman Trent Powers Chris Scully Dan Wocxidell Madison Long Candace Segar Don Miller Mike Akins Doug Bucche Ken Gabel ClitTord Marks Joe Groncki Roger Mauchline 1.
Topic: Seismic Qualification Documentation. (Joe Groncki)
Background:
RAl-ll25 asked for seismic qualification documentation for various components, l
including a large number of dampers. In some instances the damper and its accompanying motor were listed as separate items (for example 2-EB-40 for the damper and 2-EB-40M for the motor).
Parsons assumes that the damper and its motor were scismically qualified as a single unit and not as two individual pieces.
Questions:
a) Is this a correct assumption?
I Response: No. The damper andmotor were sewmically quahpedseparately.
2.
Topic: Root Cause Evaluations RP 6 (Clifford R. Marks)
Background:
During review of Condition Repor's, it was det:rmined that formal Root Cause Evaluations were performed when necessary.
PAGE 1
ADMINISTRATIVE CONFERENCE NOTES Mirch 24,1998 Questions:
a) Is there a list of all the RCE's performed between 1/1/95 till present?
Response: No list exists. An RAI will beforthcoming.
b) About how many have been performed during that time frame?
Response: About 120.
3.
Topic: RP 4, Corrective Action Program (Clifford R. Marks)
Background:
During resiew of Condition Reports, it was noticed that some forms, RP 4-1 were Resision 5.
Questions:
a) What is the latest revision of RP 4 and form RP 4 17 Response: RP 4, rev 5 is the latest revision. NU sill be sending it in response to an RAl.
4.
Topic; Containment / Enclosure Purge Supply Ductwork (Bill Clemenson)
Background:
Specification 7604-M-526 address ductuork classification for numerous MP2 duct systems including the Containment / Enclosure Building Purgc Supply. Section 6.4 of tids spec, Ductwork, does not specifically identify the Purge system but states that "All other ductuurk shall be classified as Idgh velocity ductwork." It is our assumption that the Purge supply should be SMACNA higinelocity ductwork.
Questions:
a) Please provide direction on whether this assumption is correct or not.
Response: Yes, the Contaimneni? Enclosure Building ductwork is built to S%ICNA High l'elocity Ductwork Standard.
5.
Topic: Discussion of DR-113. (NNECo)
Background:
N/A Questions:
a) N/A Response: NNECo discussed technicalissues oftheforthcoming revised response to DR-113 PAGE 2
ADMINISTRATIVE CONFERENCE NOTES Mtrch 24,1998 6.
Topic: Steam Generator Tube Rupture Analysis (Juan Cajigas) [ Continued from 3/19/98.]
Background:
We understand that the RETRAN and process data from W2-517-1015-RE Rev.1 is used in combination with dose data from xx-xxx-61RA Rev.0 (note that Rev, O has been superseded) in the "un-numbered" calculation of record because this RETRAN and process data is the same as that used in the xx-xxx-61RA Rev.0 analysis. However, inspection of the RETRAN and process data presented in xx-xxx-61RA Rev.0 reveals that this data does not match the W2-517-1015-RE Rev.1 data used in the "un-numbered" calculation.
Questions:
a) Please explain why the data can be used in combination with xx-xxx-61RA Rev.0 dose data.
Response: The "Rev. 0" dose data was obtainedfrom xx-xxx-61RA Rev.2. The "Rev.0" RETRAN data was obtainedfrom It'2-517-IO3-RE Rev.0. The updated RETRAN data was obtainedfrom ll'2-517-1015-RE Rev.2. The analysis is being re-done by Sil'EC.
Note: The above response was erroneously reported as the respcmse to Topic # 6for the 3/19/98 conference. The response to Topic #6 of3/19/98 should have been. " Deferred to 3/24/98. "
7.
Topic: PDCR 2-047-92 Rev 2 (Candace Segar) [ Continued from 3/19/98.]
Background:
PDCR 2-047-92 Rev 2 references, (in Section 5.1 under Design Inputs and Detailed Design item # 3) Main Line Engineering Report No. 89029-003 dated 2/28/90
" Technical Evaluation and Material Assessment for VOTES Hardware Installed on Motor Operated Valves in Nuclear Power Plants." The 2/28/90 date for this report corresponds to Rev 0.
Engineering Report No. 89029-003 was revised with Revision 3 being issued on 9/15/92. PDCR 2-047-92 Rev 2 shows a completed by date of 8/29/95 and approved by date of 9/11/95.
l l
Questions:
a) Were the additional revisions for Engineering Report No. 89029-003 considered dming l
Rev 2. of the modification? If yes, please identify the documentation which sb.,ws that l
Engineering Report No. 89029-003 Revs 1-3 were considered in PDCR 2-047-92 Rev 2.
1 Response: Deferred to 3'2688.
8.
Topic: Specification Revisions (Jim Collins, Wayne Dobson)
Background:
During the Tier 3 review of specifications, several specifications were noted to 1
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:
SP-GEE-48 R1 is s purchase specification containing the technical requirements for the design, fabricatio1, qualificdon testing, packaging, and delivery of electiical 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 isaued in INO. Based on the review performed the specification should not be used to ptrchase mves today since it does not appear to include all the latest technical requirements.
PAGE 3
ADMINISTRATIVE CONFERENCE NOTES Mtrch 24,1998 Specification 25203-7604-M-506 is an original Bechtel specification for the installation of containment and safeguards doctwork. 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 procedure or process prevents specifications that do not contain the latest technical requirements from being used for future work activities? [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 7]
Response: NNECo initially responded that Chapter 4 of the DCAf was the controlling document that required the design inputs to be verifiedprior to use. In Section 1.1 ofChapter Jofthe DCAf it is stated "the accuracy and limstations ofall design inputs must be determined and understoodprior to their use in the design process." NNECo then stated that Chapter 6 of the DCAf controlled the revision of specifications. Hhen asked how the use of proper specipcations was controlled for purchasing ofmaterial that was not within the scope of a design change, those presentfrom NNECo were unable to address such a condition during the procurement process. NNECofurther responded that the question was hypothetscal in nature and they knew of no instances when an outdated specyication had been used for procurement of material and that if Parsons knew of such an occurrence it should be identiped. Parsons responded that it knew of no examples where this may
?
have occurred, but the specification was listed as the latest revision in GRITS and that the question wasjust simply to identify whatprocess wouldprevent itfrcm being used. NNECo responded that the question would be answered at a later teleconference.
9.
Topie: NCR's 2-46-78 and 2-51-78 (William Keegan)
Background:
ICAVP is reviewing NCR's 2-46-78 and 2-51-78 which may have resulted in the equivalent replacement of RWST level transmitter LT-3004. NCR 2-46-78 refers to a GEMAC transmitter, NCR 2 51-78 refers to a Foxboro 12DM transmitter and PMMS lists the present l
transmitter as a Foxboro 13DM.
Questions:
a) What change document and / o installation document authorized the replacement of the transmitter from the GEMAC to the Foxboro 12DM7 And for the replacement of the Foxboro 12DM to the Foxboro 13 DM7 Response: The GEMAC transmittersfor LT-300), 3002, 3003 & 3004 were replaced with Foxtoro 13 DMper PDCR 2-284-77. The PDCR was previously transmitted to ICAl7' by RAl-549.
PAGE 4
ADMINISTRATIVE CONFERENCE NOTES M:rch 24,1998 10.
Topic: Seismic Interaction Program (Roger Mauchline) [ Continued from 3/19/98]
Background:
Document MS-35 Revision 2," Criteria and Action To Be Taken For Protection of Safety-Related Components From Damage Non-seismic Components Which Might Occur During A Scismic Event," has criteria that can be used to identify seismic interaction concerns, and also outlines the kinds of measures that can be taken when concern are identified. Seismic interaction is also addressed by the General Implementation Procedure, (GIP) which was used in the Seismic Evaluation Report. The GIP addresses scismic interaction in two ways: (1) Interactions with safety equipment (GIP item 4.5) and (2) Interactions with safety related raceways (GIP item 8.2.5).
Questions:
a) Was document MS-35 used in plant design and modifications? If so where is the documentation i
of the program or effort for this activity?
Continuation Questions.
1 Second part of Question A.
What criteria / programs were used between the original plant design and the 1986 time frame?
What scismic category 11/1 criteria was implemented to perform UIR 1769 required 11/1 reviews?
What is the status of the UIR 2567 recommendation to evaluate the need to develop a program to establish scismic 1t/1 adequacy?
Response: There have been no plant walkdowns dedicated to two-over-one (seismic interaction) concerns. These concerns were considered and avoided during plant design. Seismic interaction was considered as part of the A-46 walkdown and the Seismic Evaluation Report on this walkdown did not identip any seismic interaction concerns. Parsons Power 6fauchline) noted that the A-46 walkdown only calledfor seismic interaction inspection in the vicinity of equipment and electrical raceways.
Regarding the status of UIR 2567, NNEco p~aphrased the conclusion ofA'R #9700981101. This AR, uhich was the response to UIR 2567, said that it was not necessary to have a two-over-one program because these concerns were addressed during plant design.
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i ADMINISTRATIVE CONFERENCE NOTES M rch 26,1998 DATE:
3/26/98 PURPOSE:
Administr ative telephone conference with NNECo, NRC, NEAC and Parsons to discuss:
1.
RWST level indicators 2.
PDCR 2-5-77 3.
PDCR 2-047-92 Rev 2 4.
Emironmental Qualification Schedule
- 5. Surge Suppressers for MOV SV-4188
)
6.
8" HBD-196 AFW Turbine Exhaust Line Scismic Adequacy j
LIST OF ATTENDEES:
NNECo NRC NEAC Panons Joe Fougere Ralph Architzel Wayne Dobson Fred Mattioli Don Marks Ken Fox Keym lepson Kalvin Anglin Jack Lawton Steve Unikewicz Ken Gabel Andres Malias Kent Russell Roy Terry Dale Pmitt N bil Juraydini Larry Wigley Rich Ewing Mike Akins l
i 1.
Topic: RWST level indicators (Kent Russell) l
Background:
Walkdown indicates that RWST level indicators Ll3001/2/3/4 have been replaced with indicators different than original design, yet we have found no documentation of their replacement.
Questions:
a) llave RWST level indicators Ll3001/2/3/4 been replaced?
If so, please identify the documentation that replaced the indicators.
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Response: Deferred to 3/3V98 l
l 2.
Topic: PDCR 2-5-77 (Jack Lawton)
Background:
This modification, which was canceled, would have installed a diode in series with the ESS scal-in contact (see drawing 32041, sheet 6) in the diesel generator auto start circuit. However, the drawing shows a diode, in parallel with a capacitor, installed in this circuit.
Questions:
a) If this reficcts actual neld conditions, and is not part of the original design, what document was used to make the design change and installation?
Response: The diodes were installed in April,1975, and verified on red-hned Bechtel drawings.
7he modificction was developed using NE drawings, which did not show the diodes. IIhen it was realized the diodes existed, the modification was canceled PAGE1 I
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ADMINISTRATIVE CONFERENCE NOTES M:rch 26,1998 3.
Topic: PDCR 2-047-92 Rev 2 (Candace Segar) IContinued from 3/19 & 24/98.]
Hackground: PDCR 2-047-92 Rev 2 references, (in Section 5.1 under Design inputs and Detailed Design item # 3) Main Line Engineering Report No. 89029-003 dated 2/28/90
" Technical Evaluation and Material Assessment for VOTES Hardware Installed on Motor Operated Valves in Nuclear Power Plants." 1he 2/28190 date for this report corresponds to Rev 0.
Engineering Report No. 89029-003 was revised with Revision 3 being issued on 9/15/92. PDCR 2-047-92 Rev 2 shows a completed by date of 8/29/95 and approved by date of 9/11/95.
I Questions:
a) Were the additional revisions for Engineering Report No. 89029-003 considered during Rev 2. of the modification? If yes, please identify the documentation which shows that Engineering Report No. 89029-003 Revs 1-3 were considered in PDCR 2-047-92 Rev 2.
Response: The later revisions of the engineering report did not afect the modification, and were not considered.
4.
Topic: Emironmental Qualification Schedule (Mike Akins)
Background:
RAl 1105 requested specific information related to the Emironmental Qualification j
j Program. In a response prosided by NNECo, a revised EEQ Program schedule was provided and received on March 11,1998.
Questions:
a) In relation to the resised schedule to completing EQRs provided as part of this response, please discuss the impact to the overall Unit 2 ICAVP schedule.
Response: The overall schedule currently ends on 11/30'98. NNECo stated that they are currently re-evaluating the EQ schedule and it is their intent to accelerate thte schedule to complete by 9/15/98. To accomplish this acceleration, NNECo has contracted DESI to acquire new EQ resources. Parsons questioned th realistic availability ofthese resources. NNECo stated that DES 1 would advise them. 7he NRC noted that this was a new date and was beyond what they previously understood.
Environmental Specification completion may slip due to problems completing IIELB analyses and the performance of additional radiation assessment calculations. This is seen as the most significant obstacle at present. 7 hey have credited new harsh environments and chang 3ed radiation only to steam and radiation.
The EQML is on schedule. They have added 20-30 components to the list. More may be added as the environmental specification isfinah:ed.
NNECo noted that 77bts have been completed. (they also mentioned draft EQRs. but this may have been an error.)
PAGE 2
F ADMINISTRATIVE CONFERENCE NOTES Mrrch 26,1998 5.
Topic: Surge Suppressers for MOV SV-4188 (Dale Pruitt)
Background:
PDCR 2-89-107, (Surge Suppressers for MO', SV-4188), stated that a walkdown discovered 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 32012, Sheet 49. The modification data base indicates the modification was canceled.
A search of the PMMS Work History Database did not locate a work request that addressed the condition.
Questions:
a) Was this Modification canceled?
Response: s es.
b) If yes, are the surge suppressers installed?
Response: Three out ofthefive shown were installed. A search will be conducted tojust@ this condition or a condition report will be written. Action to be given at next conference.
c) If yes, what permitted their installation?
Response: Deferred until next meeting.
6.
Topic: 8" HBD-196 AFW Turbine Exhaust Line Seismic Adequacy (Ken Gabel) l
Background:
Engineering Record Correspondence ERC-25203-ER-97-014 documents an evaluation of the seismic qualification requirements for the 8"-HBD-196 Terry Turbine exhaust piping. These evaluations included installation walkdowns as part of MP2 assessments in accordance with C-L 81-14 and GL 87-02 (SQUG/ GIP). The GIP assessment portion focused on pipe span lengths and the existence of rod hung supports. The GIP Program addresses qualification techniques for 22 difTerent classes of equipment. Passive equipment, which includes piping, is not included within these 22 classes. In NNECo's 9/21/92 GL 87-02 response letter to the NRC, a commitment i
was made that any deviation from the GIP guidance would be justified, documented, and communicated to the staff.
Questions:
a) What was the rationale and justification for use of the GIP Program?
b) What GIP caveats, inclusion rules, similarity requirements were looked at?
c) What is the basis for considering a variable spring support as resisting sakmic effects?
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ADMINISTRATIVE CONFERENCE NOTES l
Mirch 26,1998 l
- d) Where any other programs used, i.e., EPRI Report NP-6041, SSRAP Reports?
c)' Has the GIP Program been used/ referenced elsewhere for qualifications outside of its intended I
bounds?
Response: ERC-25203-ER-97-014 summari:es the eforts to establish seismic adequacy, not 8
seismic quahfication for the 8"-HBD-196 Terry Turbine exhaust piping. The objective was to i
substantiate the structural integrity to show that there would not be seismic interaction of the exhaust i
piping with any other seismic category l equipmenttinstallations in the area. The GL 81-14 served as the basisfor the seismic adequacy positionjustJ1 cation. SQUG/GIF program was usedfor turbine driven pump qualification. It was indicated that there was no knowledge of application of SQUG/G1P outside ofits intended bounds.
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ADMINISTRATIVE CONFERENCE NOTES March 31,1998 DATE:
3/31/98 PURPOSE:
Administrative telephone conference with NNECo, NRC, NEAC and Parsons to discuss:
- 1. Alternate Shutdown Panel
- 3. Status of EWR's included in the Tier 3 review sample 4.
NNECo response to DR-0160 l
- 5. Specification Revisions 6.
Discussion of DR-127 7.
Discussion of DR-119 LIST OF ATTENDEES:
NNECo NRC NEAC Parsons Joe Fougere Steve Reynolds Wayne Dobson Fred Mattioli Don Marks Ken Fox Jim Giova l
Ken Moore Jim Collins Greg Tardif Ray Thomas John Lockerby Kent Russell Dan llunlev Andy O'Connor l
Dan Van Duvne Dale Pruitt George Pitman Ken Mayers Cris Cristallo Lany Wigley Dave Bajumpa Bob Steinmetz Mike Akins l
Joe Groncki 1.
Topic: Atternate 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 condd not befound regarding the testing of the Alternate shutdown Panel (C-21).
Individualcomponents do have surveillance proceduresinplace See Tech Spec Table 4-3-6:
Comr>onent Procedure Channel Check SP-2619 E Channel Calibration (Reactor ColdLeg Temperature)
No surveillance procedure (Note: a CR has been generated, a number to be supplied.)
PAGE1
/
ADMINISTRATIVE CONFERENCE NOTES Mrrch 31,1998 Response Continued:
Cbmponent Procedure i
l Pressurizer Pressure Low range SP-2402.,
)
High range SP-2402 B 1
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Pressuri:er Level SP-2402 E l
Steam Generator Level SP-2402 D Steam Generator Pressure SP-2402 C l
l 2.
Topic: Surge Suppressers for MOV SV 4188. Follow on questions to 3/26/97 Conference l
(Dale Pruitt) l
Background:
PDCR 2-89-107 (Surge Suppressers for MOV SV 4188) identified that during l
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 i
modification data base indicates the modification was canceled.
A search of the PMMS Database did not locate a work request that addressed the condition.
l Per the conference of 3-26-97 three of the five suppressers were found in the field l
Questions:
l a) What qualified the installed suppressers?
Response: Drawing 800236D i
b) What installed the suppressers since no work doce:nent could be found?
t i
Response: PDCR 2-140-81 Note: the PDCR and associated A IVO will be requested on an RA1.
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, i
l DCN etc. ) for three remaining EWR's.
Questions: Please provide status and design change documentation citations for:
a) EWR # 2-97-007, Thermo-Lag Fire Barrier Response: EliR is open and no design change is assigned.
PAGE 2
ADMINISTRATIVE CONFERENCE NOTES Mrrch 31,1998 b) EWR # 2-96-156, Condensate Min. Flow Recirc Response: EliR is open and DCR # Af2 97-520 was assigned.
c) EWR # 2-92-172, RBCCW Rupture Disc ( Note: RAI-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 ofEllR # 2-92-172, RBCCll' rupture disc EliR is still open and DCR H Af2 018 was assigned. Per the results of the conference call, NNEco was to submit the DCR on the next day, however, this was not received. It'ayne will discuss this with NNECo during the status meeting on Tuesday (4/7/98).
4.
Topic: NNECo response to DR-0160 (Jim Giova)
Background:
PDCR 2-45-82 changed the thermal margin / low pressure (TMILP) 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 vendor manuals are:
VTM2-150-007A," Bistable & Auxiliary Trip Unit", Rev. O, 12-04-87.
VTM2-150-008A, " Reactor Protection System (Vol.1)" Rev. 01, 2-11-86.
VfM2-150-009A, " 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 term, the 75 psia, or the char.ge in the resistor ehm 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 control document that was used to update VTM2-150-008A. GRITS indicates that there are no change control documents posted against the VTM.
Questions:
a) When and how was VTM2-150-007A updated to reflect this change?
Response: NNECo stated that they cannot currertly locate the vehicle used to update 17Af-150-008A. NNECo said they wouldforward additional information to support their disposition of DR-0160 regarding the change to 1 TAf-150-008A.
Parsons asked how we can concur with NNECo 's response to DR-0160 that i TAfs -150-007A and 009A were not impacted by the revision to the Thermal Afargin Pre-trip Ojpet term or the change of two resistors' chm rating. Parsons suggested writing a RAI to obtain copies of the vendor technical manuals QTAf), ifthey are not too voluminous. NNECo stated that they are reluctant toforward the iTAfs 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.
PAGE 3
ADMINISTRATIVE CONFERENCE NOTES Mrrch 31,998 5.
Topic: Spe.:ification 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:
SP-GEE-48 R1 is a purchase specification containing the technical requirements for the design, fabrication, qualification testing, packaging, and delivery cf 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 r ' 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 technical requirements from being used for NON-MODIFICATION work activities, such l
as 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 l
review confirms it meets the latest requirements?
Response: NNECo responded that the NGPs associated with purchase orders, replacement item evaluations, and commercialgrade dedication requires that the purchase request be reviewed by l
engineering to ensure the correct EQ, seismic, and other technical requirements are properly l
Imposed. Examples in procedures NUCMPM 3.01, NGP 6.02, and NGP 6.12 were given. A review of NGP 6.02, Quality Material Requests and Quality Purchase Orders, confirmed that responsibilities and requirements are contained in sections 5.3, 5.5, 5.6, 6.1.1.3, 6.1.1.6, 6.1.3.1, 6.1.3.3, and l
6.1.3.4. A review ofNGP 6.12 Evaluation ofa Replacement item, confirmed that responsibilities and l
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) l Response: NNECo presented their revised response to DR-127.
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o ADMINISTRATIVE CONFERENCE NOTES M:rch 31,1998 i
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- 7. Topic: Discussion of DR-Il9 (NNECo Requested Topic) i Response: NNECo presented their revised response to DR-119.
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