ML20199D353
| ML20199D353 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/18/1997 |
| From: | Schopfer D SARGENT & LUNDY, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 9583-100, NUDOCS 9711200317 | |
| Download: ML20199D353 (53) | |
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' 312 269 6018 November 18,1997 Project No. 9583-100 t
Docket No. 50-423 Northeast Nuclear Energy Company Millstone Nuclear Power Station, Unit No. 3 Indepe. Jent Corrective Action Verification Program [
United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 I have enclosed the following twenty-one (21) discrepancy reports (DRs) identified during our review activities for the ICAVP These DRs are being distributed in accordance with the Communications Prote.ol, PI MP3-01.
DR No. DR MP3-0177 DR No. DR VP3-0597 DR No. DR-MP3-0293 DR No. DR MP3-0600 DR No. DR-MP3-0384 DR No. DR-MP3 0613 DR No. DR MP3-0397 DR No. DR MP3-0621 >
DR No. DR MP3-0438 DR No. DR-MP3-0624 DR No. DR-MP3-0455 DR No. DR-MP3 0631 ,
DR No. DR-MP3-0473 DR No. DR-MP3-0641 j DR No. DR-MP3-0495 DR No, DR MP3-0570 DR No. DR-MP3-0643 DR No. DR-MP3-0649 /
jy -
j DR No. DR-MP3-0572 DR No. DR MP3-0650 -
DR No. DR-MP3-0652 I have also enclosed the following one (1) DR that has been determined invalid. No action is required from Northeast Utilities for this DR. The basis for its invalid determination is included on the document.
DR No. DR-MP3-0510 9711200317 971118
^ "
E" " E' 11111Illilli Ul!! Bill!
. 55 East Montce Street
- Chicago, IL 60603 5780 USA
- 312 269-2000
(
United States Nuclear Regulatory Commission November 18,1997 Document Control Desk Project No. 9583 100 Page 2 I have also enclosed the following three (3) DRs for which the NU resolutions have been reviewed and accepted by S&L.
DR No. DR-MP3-0162 DR No. DR-MP3-0329 DR No. DR-MP3-0470 I have also enclosed the two (2) DRs for which the NU resolutions have been reviewed but not accepted. S&L comments on these resolutions have been provided.
DR No. DR MP3-0180 DR No. DR MP3-0204 Pleue direct any questions to me at (312) 269-6078.
Yours very truly, k.N. C D. K. Schopfer Vice President and ICAVP Afanager DKS:spr Enclosures Copies:
E. Imbro (1/1) Deputy Director, ICAVP Oversight T. Concannon (1/l) Nuclear Energy Advisory Council J. Fougere (1/l) NU m Waghwt\97svill8-a &c l
Northeast UtilNies ICAVP DR No. DR.MP3 0177 Millstone unit 3 Discrepancy Report Rev6ew Oroup: system DR VALID I ' 9' Potential Operoidisty lasue 06ecapline: Meelexal Doogn D6ectopancy Type: Component Date O va No SysterWProcess: Rss NRC sWance level:4 Date faxed to NO:
Date Put4shed: 11/2097 Diar *Paacy: FSA5~Sec 6.3.2.2.5 reqmt for gate valves w/ flexible wedges is inconsistent w/ spec 2282.050-676.
DescHpt6on: FSAR Section 6.3.2.2.5 states that the seating design for all emergency core cooling system motor operated gate valves are of the Crane flexible wedge design. Containment recirculation system components are described in FSAR Section 6.3 to be included in the emergency core coooling system. Valve specification 2282.050-676 through Revision 1 shows the containment recirculation spray system valites 3RSS*MOV38A,B having a solid wedge design.
Review Valid invaled Needed Date Instsstor: Feingold. D J. O O O $1/11'S7 VT Lead: Nort, Anthony A O O O 15'10S7 VT Mgt: schopfer, Don K G O O 55'S7 IRC Chnm: Singh, Anand K O O O i17/S7 Date:
INVALID:
Date:
RESOLUTION:
Preyhusly identifled by NUF O Yes 9) No Non D6ecrepent Condet60n?Q Yes #1 No Resolution Pending?O ve. *) No Resolut6on Unresolved?O ve. *) No Review in6t6ator: Feingold, D. J VT Lead: Nort. Anthony A VT Mgt: schopfer, Don K O O O IRC Chmn: Sngh. Anand K D.t. .
sL Comments:
Pnnted 11/1&97 2.2511 PM Page 1 of 1
4 Northeast Utiinies ICAVP DR No. DR MP3 0293 l
Ministone unk 3 Discrepancy Report l Rev6ew Group: System DR VAUD Review Element: System Ongn !
pg y, 06ecipune: Mediamcal Deegn D6ecrepancy Type: Compone d Data O Ya 7,, g, SystemProcoes: RSS NRC S:g:Tme level: 4 Date faxed to NU:
Date Published: 11/2397 D6.crepancy: FSAR Table 6.2 62 is inconsistent with specification 2214.602-044 & drwg 2214.802 044 018.
Ducript6on: FSAR Table 6.2 62 states that the containment recirculation pumps are fabricated of ASTM-A452 Type 304 stainless steel.
However, containment recirculation pump design specification 2214.802 044 through Revision 1 and drawing 2214.802 044-018 Revision G show no pump parts made from ASTM A452 Type 304 stainless steel.
Review Valid Invel6d Needed Date initiator: Feingoki, D. J. O O O $ 5' 87 VT Lead: Nort Anthony A G O O 1$ 0S7 VT Mgr. Schopfw, Don K Q Q Q 11/1497 IRC Chmn: Snngh, Anand K Q Q Q 11/1747 Dele:
INVALID:
Date:
RESOLUTION:
Previously ides Whed by NU7 O Yes Lei No Non Discrepant Condition?Q Yes '#) No Resol 6'60n Pending7C# Yes M No RuoluHon Unresolved?O Ya + No Review Acceptab!s Not Acceptable Needed Date VT Lead: Nort, Anthony A VT Mgr: Schopfer, Don K IRC Chmn: Singh, Anand K Date:
SL Comnents:
~
Pnnled 11/1&912348 PM Page 1 of 1
l o !
I Northeast Utilities ICAVP DR No. DR MP34344 Millstone Unit 3 Discrepancy Report ,
Review Group: Accxiont Magebon DR VALID Review Element: Operehno Procedure J 06ecipl6ne: Otho' O yes 1 Diecrepency Type: Uconsang Documert pg
~
SystesWProcess: WA NNC Sign 6Aconce level: 3 Date faxed to NU: 1 Date Putdished: 11/2o97 D6ecrepancy: Verification of time critical operator actions following a CVCS line bresk i
Description:
FSAR Section 15.6.2 evaluates the consequences of the worst case line break outside containment, which is a rupture of the letdown line. The FSAR contains the statement:
" Area radiation and leakage detection instrumentation provide the means for detection of a letdown line rupture. Frequent operation of the CHS reactor make!Jp control system and other CHS instrumentation also aids the operator in diagnosing a letdown line rupture."
In addition, in the Safety Evaluation Report (SER) for Millstone 3, the NRC states that the operator can respond to a change in '
the level in the Volume Control Tank (VCT) to identify and isolate the break.
From the above one can conclude that for a letdown line rupture outside containment the operator can rely on:
- 1. Radiation detection Instrumentation
- 2. Leak detection instrumentation
- 3. VCT makeup control and instrumentation The ICAVP reviewed the following documents in order to verify that there are area radiation monitors and level and leakage detection capabilitias sufficient to detect and mitigate the design basis CVCS leak of 152 gpm within 30 minutes:
A. EOP 35 ECA 1.2 LOCA Outside Containmer,t.
B. Unit 3 Millstone FSAR Section 15.6.2.
C. AOP 3555 Rev,7 Reactor Coolant Leak.
D. 0;.erating Procedure 3353.MB3A Rev.1 VCT Level Hl/LO.
E. NUREG 1031, Safety Evaluation Report, Section 15.6.2 F. EOP 35 E 0 Reactor Trip or Safety Inkction G. EOP 35 E 1 Loss of Reactor or Secondary Coolant The FSAR on page 15.6-3 states that the operator has diverse means of detecting the letdown line rupture through:
- a. Area radiation monitors.
h I onlr rimiarfinn incintrnantntinn PrWed 11/1&97 2.27.32 PM Page 1 of 3 ;
Northeast UtiDties ICAVP DR No. DR MP3-0384 Millstone Unit 3 Discrepancy Report
- c. Frequent operation of the OHS reactor makeup control system and the CHS instrumentation.
If the break is large enough, the reactor will trip and the and the operator will enter the Emergency Operating Procedures (EOP's) for reactor trip and LOCA (Documents F and G). In both of these procedures, a high alarm on an area radiation monitor outside containment will cause the operator to enter the EOP for the LOCA outside containment (Document A). Note that the only entry point into this EOP for a LOCA outside containment is an hbnormal radiation reading. In the event of abnormal radiation in the auxiliary building, explicit directions are given to close the outer and inner containment isolation valves,3CHS*CV8152 and 3CHS*CV8160, respectively.
To determine if there is another means for the operator to isolate the letdown line rupture using other instrumentation, additional procedures were reviewed.
The stated purpose of the Abnormal Operating Procedure (AOP) for reactor coolant leakage (Document C) is to provide the actions r.ecessary for evaluating the magnitude of a small(within the car,acity of the Charging System) Reactor Coolant System leak in Modes 1,2,3 or 4. Some of the entry conditions that may indicate a CVCS leak outside containment are:
- a. Pressurizer level decreasing slowly,
- b. Unexpected increase in charging flow,
- c. Makeup volume increasing or abnormally h;gh.
- d. RCS inventory computer program indicating an abnormal leak rate.
Following the directions of the AOP leads an operator to determine the leak rate using the Volume Control Tank (VCT) level trend. The procedure then directs the operator to proceed to try to identify the location of the leak. At no time in this AOP is the CVCS letdown line rupture stated as a possible cause of the leakage or are directions given to verify that 3CHS*CV8152 and/or 3CHS*CV8160, the containment outer and inner letdown isolation valves, are closed.
As noted above, the NRC stated in the SER (Document E) that the VCT level indicator wow be used to identify and isolate the letdowr line rupture. The operating procedure for response to volume control tank level alarms (Document D) does not identify:
i) A possible cause of low VCT level is a letdcwn line rupture as stated in the SER -
li) The valve numbers for the containment letdnwn isolation '
lii) Operator action to isolate letdown within 30 minutes of a design basis CVCS leak of 152 gpm CQ 'CLUSION in summary, abnormal radiation readings outside containment coincident with reactor trip or loss of coolant will invoke the PrWed 1 inst 7 2 27.36 PM Page 2 of 3
ICAVP DR No. DR44P3-0344 Northeast Utilities Millstone unit 3 Discrepancy Report EOP's which will lead to letdown isolation. There is no indication in the procedures reviewed that RCS leakage without a radiation monitor reading, or a change in VCT level, will lead to the isolation of the letdown line. Furthermore, the operator is not provided guidance to isolate letdown of a design basis CVCS leakage of 152 Opm within 30 minutes, Based on this review, there is a discrepancy between the existing operating proceduros and the commitments in the FSAR and SER.
Rev6ew Valid loveled Needed Date initiator: Schwartz, Berry Q Q Q 1o01/97 VT Lead: Rehop, Raj D Q f' O 10/31/97 VT Mgri Schopfer. Don K O L O 15'S'87 IRC Chmn: Cin0h, Atiend K Q Q Q 11/17/97 Dei.:
INVAllo:
Sete:
RESOLUTION:
Prev 6ously identifled by NUP C) Yes iGl No Non D6ecropont Condet6on?C) Yee 19) No
~
Resolut6on Pending?O vee ' * > No ResoluGon Unresolved?O Yee 4) No Review Acceptable Not Acceptable Needed Date g,g; VT Leed: Rehop, Raj D VT Mer: Schopfer, Don K I 1RC Chmn: Singh, Anand K
- e:
SL Comments:
Printed 11/1897 2.27.39 PM Page 3 of 3 l
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Northeast Utilities ICAVP DR No. DR-MP3 0397 uisistone Unit 3 Discrepancy Report Review Geoup; syelem DR VALID Rwww tw syworn Onion p,,,,,,,, op,,,,,,, ,,,,
Diecipline: Mechenecal Dagn O vee D6ecropsney Type: Calculeton ;8) No System 9toceae: $WP NRC tign6Acence level: 4 page faxed to NU:
Dele Published: 11/2097 D6 crepency: Calculation 95 ENC.1177 M3 rev. O, CCN 1, has several unjustified values.
Ducr$t6on: Review of calculation 95-ENG 1177 M3 Rev. O through CCN 01 identified the following discrepancies:
1.) The calculation uses design inputs which are documented inclusively in calculation 90-0691130-M3. However, Design input #6 of the calculation (95-ENG 1177) references a portion of calculation 90-069 't 130 M3 wt lch has been superseded by calculation 95-ENG 1177 M3. Therefore, no valid reference for design input #6 is documented in the calculation.
2.) Design input #12 is based on Reference 6. However, the number of passes in the heat exchanger is not contained in Reference 6.
,,4 Assumption 1 of the calculation,'no subcooling in the coi densor", contradicts the 8 'F of subcooling assumed by the vr idor suppling the heat exchanger data as documented in 6eference 6 Attachment 1, Itcm 7.
4.) No reference or justification for the Correction Factor, F=1.0 for the heat exchanger is documented in the calculation.
These discrepancies are not expected te have a significant effect on the calculational conclusions.
Review Valid invaled Needed Date initiator: Dionne, B. J. O O O 1 2S7 VT Lead: Neri, Anthony A O O O 11/12S7 VT Mgt: Schopfer Don K O O O '5/i'S7 1RC Chmn: singh, Anand K O O O 11'17/S7 Dm.:
INVA8 lD:
Date:
RESOLUTION:
Previously identif4d by NU7 O Yee 9) No Non D6screput Condition?Q Yee it) No Resolution Pending70 Yee '5) No Aesolut6on Unresolved?O vee <s) No Review g Acceptable Not Acceptable Needed Date VT Lead: Nort, Anthony A vi u r:v schopter, Don K O Q O 1RC Chmn: singh, Anand K Printed 11/1&S7 2 29 s3 PM Page 1 of 7
. DR N3. DR-MP3 0397 Norttwast UtilNies ICAVP '
Millstone Unit 3 Discrepancy Report
- . . . . . . . . , - ~ , .
O O O SL Comments: !
d Prned 11/1897 229 57 PM Pope 2 of 2
_ _ - - - - - _ _ _ _ _ ~ - . _ . , - _ . . - ~ . _. . . .
0 Northeast Utilities ICAVP DR No. DR MP3-0434 Millstone Unit 3 Discrepancy Report
$evtew oroup: SyWom DRVAUD I *I
- Potential operebility issue te L.,.: Mutenal Demon O vee Diecrepency Type: Cornponart Date e) No Systemfroceae: CSS NRC segrdhcance level: 4 Dale faxed to NU:
Date Putdiohed: 11CQ97 06*cr*Pency DBSD 3DBS-NSS-002 is inconsistent with drwg 2214.602 040-012 with respect to OSS pump NPSHr.
D+*cript6on: Design Basis Summary Document 3 DOS-NSS-002 Revision 0, Section 12.2.3, states that the net positive sucihn head required for the quench spray pump is 23.8 feet at the tr,aximum expected runout flow of 5,000 gpm. Design drawing 2214.602-040-012 Revision A shows that the net positive suction head rt.luired by the quench spray pumps is approximately 16 feet at 5,000 gpm.
Calculation P(R) 1062 Revision 0 shows that the net positive suction available to the quench spray pumps at a pump flow rate of 5,000 gpm is 23.8 feet. Using the.Vrlue of 16 feet for required net positive suction, yields a net positive suction head margin at 5,000 ppm of 7.8 feet. Hwever, the Design Basis Summary Document requirement 11 inconsistent with design drawing 2214.602 040-012.
Review Vehd inval6d Needed CWe init6stor: Feingo6d D. J. O O O 1 t'1 $'S7 VT Lead: Nort, Anthony A @ Q Q 11/11/97 VT Mgri Schopfer. Don K O O O itS7 IRC Chmn: Singh, Anend K O O O 15'17/S7 Deie:
INVALID:
Date:
REGOLUTION:
Previously identiflod by Nu? O Yes F No Non Descrepent Condet6on?O ves i No Resolution Pend 6ng70 Yee 6) No Renoiuiion unre.oeved70 veo + No Review Accepteble Not Acceptable Needed Date VT Lead: Nort, Anthony A VT Mgr: Schopfer, Don K IRC Chmn: Singh, Anend K Date:
St Commentat., ,
PrtrWed 11/1&97 2.35:12 PM P.ge 1 of 1
, . . ~ _ . .
I l
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Northeast Utilities ICAVP DR No DR MP3 0486 j Millstone Unit 3 Discrepancy Report Rev6ew Oroup: Synkrn DR VALIO Review Element: tystem Desyn p g ,
D6ecipline Mechancel Des'"
D6ecropency Type: Licensing Document O va g
Systemfrocess: N/A NRC SW level: 4 Date faxed to NU:
Date Published: 11/2G97 Okcr*Pency: CDA Setpoint in FSAR Section 6.2.2.2, FSAR Table 6.2 5, and FSAR Table 6.216 D**crtPtion: The Technical Specifications and the FSAR list different values for the Containment Depressurization Actuation (CDA) setpoint.
Technical Specification (TS) Table 3.3-4 states that the CDA setpoint is 8.0 psig and shall not exceed 8.8 psig.
FSAR Gection 6.2.2.2
- Containment Heat Removal System, System Design" states that the CDA setpoint is 23.5 psia (8.8 psig). This contradicts the CDA setpoint of 8.0 psig in TS Tablo 3.3 4.
FSAR Table 6.2 5, " Accident Chronology, Hot Leg DER," states that the CDA setpoint is 10.0 psig. This is based on a conservative assumption in the containment pressurization analysis (Table 4 of US(B) 273, Rev. 5), but the FSAR table does not identify why it contradicts the CDA setpoint of 8.0 psig in TS Table 3.3 4.
FSAR Table 6.216 " Accident Chronology, Pump Suction DER with Minimum ESF," states that the CDA setpoint is 10.0 psig.
This is based on a conservative assumption in the containment pressurization analysis (Table 5 of US(B) 273, Rev. 5), but the FSAR table does ; nt identify why it contradicts the CDA setpoint of 8.0 psig in TS Table 3.3 4.
These differing values for the CDA setpoint in the Technical specification and the FSAR is a discrepant condition.
This discrepancy applies to both the QSS and RSS systems.
Review Valid invalid Nooded Date inNietor: Wakeland, J. F, 8 O O 11'12S7 VT Lead: Nerl, Anthony A O O O 11/12.S7 VT Mgr: Schopfer. Don K O O O 11'17'87 IRC Chmn: singh. Anand K O O O 11/17 S7 D.ie:
INVALID:
Date:
RESOLUTION Prevlously identifled by NU? O vos ei No Non Discrepant Condalon?O vos #1 No Resolut60n Ponding70 ve. + No Resolutbn Unresolved?O vos
- No Review Pnnted 11/16-97 2.3e 23 PM Page 1 of 2 l
l - _ .
4 Northeast Utilhies ICAVP DR N3. DR-MP34456 Millstone unN 3 Discrepancy Report Initiator: (none) e'm - sae me- .w u.
VT Lead: Nort, Anthony A O O O VT Mer: Schopfw, De K mc chmn: s v .An.m x =
Date:
SL Commets:
l Prtnted 11/1&97 2.38.27 PM Page 2 g 2
Northeast Utilities ICAVP DR N3. DR MP3 0473 Mimone Unit 3 Discrepancy Report Revtow Group: System DR VALID I N'* Potent 6al Operoidity leeue D6ec6pune: Ce*'
D6ecropency Type: Calculation O vos 9, g, SystemProcess: SWP NRC &lgnencence level: 3 Date faxed to NU:
Dale Published: 11/2097 Dhwy: Closing the wrong SWP valve to tenninete a break in line 3 SWP 750 245 3 D**cr6P46on: To isolate a break in line 3 SWP 750 245 3, Calculation 12179-P(R) 1198, Rev. O, ' Flood Source Termination Evaluation,'
states that either valve 3SWP*V835 or valve 3SWP*V834 should be Closed (page 13), A review of Drawing 25212 26933 sheet 2 Rev. 34 concludes that valve 3SWP*V836 should be closed instead of 3SWP*V834 and both valves 3SWP'836 and 3SWP'835 shold be closed to mitigate the break.
There are no AOPs or EOPs associated with this action.
Rev6ew Vend invoied Needed Date Ordt6etor: Leunt C. M. O O O 15'5S7 VT Lead: Nwt, Anthony A B O O i'5S7 VT Mgr: Schopfw, Don K O O O ii'1SS7 1RC Chmn: singh, Anand K G O O $ $' 7'S7 Dm.:
INVALID:
Date:
RESOLUTION:
Prov6ously ident6Aed by NU? O vos 97 No Non Discrepent Condition?O vos *) No Resolution Pending?O ve. *) No Resolut6on Unresolved?O vos *) No Review Acceptable Not Acceptable Needed Date VT Lead: Nort, Anthony A VT Mgt: Schopfer, Don K IRC Chmn: Singh, Anand K D.e sL Carmente:
PrWed 11/1597 2 39 Oo PM Page 1 of 1
Northeast Utilities ICAVP DR No. DR MP34496 Millstone Unit 3 Discrepancy Report Rev6ew oroup: syotem DR VALID
- Potential Operetnisty lasue D6ecip46ne: Mechancel Desgn D6ecrepancy lype: congene's Date Om it) No systemProcess: Oss NRC Significance level: 4 Date Faxed to NU:
Dele Putdished: 11/2097 6ecropency: Inconsistencies between valve design and FSAR Table 6.1 1 with respect valve stem materials.
Deecript6on: FSAR Table 6,1 1 states that valves employed for engineered safety features shall have stems made of material specification SA 182 F316 or SA 187 F304. In contrast, the following valves have stems made of material specification A 276 Types 304 and 016 as shown be'v Component Stem Material Reference CSS Mnl Vivs A 276 drwg 2282.150154 021 Revision G 30SS*V50 58 TP 316 Cond B 30SS*V940, V941 3OSS*V948 V951 3OSS*V968 V975 QSS Mnl Vivs A 276 drwg 2282.150-154-043 Revision G 30SS*V933 V936 TP 316 Cond B 30SS*V957 V959 3QSS*V964-V967 3QSS*V986, V987 QSS Mnl Vivs A 276 drwg 2282.050153-039 Revision B 30SS*V42, V4J TP 304 Cond A The valve materials in FSAR Table 6.1 1 represent gate and globe valves. Butterfly valves and check valves do not have valve stems. Therefore, check valves and butterfly valves are not represented by the valve components specified. However, the FSAR Table 6.1 1 does not identify the type (s) of valves represented by the materiallist. If the stem materiallisted in FSAR Table 6.1 1 is meant to apply to butterfly valve shafts, other stem materiel dix.repsacles would exist.
Review Val 6d invetid Needed Date initietor: Feingold, D. J O O O 11'10S7 VT Leed: Nerl,RAhonyA O O O 11/15/97 VT Mgt: schopfer, Don K Q
@ Q 11/1497 1RC Chmn: Singh, Anand K O O O 5'/27'S7 ernea nissiisioTi"~ ~~~~~----~ es,. i or 2
Northeast Utilities ICAVP DR N2. DR-M33-0495 Millstone Unit 3 Discrepancy Report u .:
IPNALID:
Date:
RESOLUTION:
Prev 6ously identeSed by NU? U Yes 9) No Ncn Discrepent Condition?Q Yes '91 No Resolut60n Pending?O Yes I No Resolution Unresobed?O Ye. @ No Rev6ew Acceptable Not Acceptable Needed Date VT Lead: Nort, Anthony A -
VT Mgr: Schopfer Don K 1RC Chmn: Singh, /.nend K Date:
SL Conwnents:
t 1
1 Printed 11/18S7 239 44 PM Page 2 of 2
' I Northeast Utilities ICAVP DR N . DR-MP3-0670 Millstone Unit 3 Discrepancy Report Review Group: Jystem DR VALID ;
Potential Operability issue D6ecipi6ne: Mechanical De**"
D6ecrepency Type: Component Deta O Ya f5) No
~
SystenVProcess: RSS NRC SW level: 4 Date faxed to NU:
Date Published: 11/2G97 I D6*cr*Pency: Design specification (s) & drawings not available for RSS pumpi I seal head tank valves.
Ducription: P&lD EM-112C Revision 18 and the plant computer data base, PDDS, show the following val"es for the containment recircutstion pumps seal head tanks, but no vendor drawing or specification is identifed for these valves:
3RSS*V38 through 3RSS*V45 3RSS*V50 through 3RSS*V53 P&lD EM-112C in conjunction with PalD Legend EM 1008 Revision 10 show the above listed valves to be globe valves.
The PDDS references specification 2472.800-943 Revision 10 for these valves. However, this specification does not address globe valves, only check needle, and ball valves.
Review valid invalid Needed Date initiator: Femgold. D. J- G O O 11/10/87 VT Leed: Nort, Anthony A g [ [ 11/11/97 VT Mgt: Schopfer, Don K G O O 11/17/S7 IRC Chmn: Singh, Anand K G O O 11/17/87 Det.:
INVALID:
Date:
RESOLUTION:
Previously identined by NU7 Q Yes el No Non Discrepent CondetiontO Yes 4 No Resolution Pending?O Yes (6) No Resoiuisonunresoeved70 Yes 5) No Review Acceptable Not Acceptable Needed Date VT Lead: Neri, Anthony A VT Mgt: Schopfer, Don K IRC Chmn: Singh. Anand K Date:
SL Comments:
9 Printed 11/1&97 2.4o:36 PM Page 1 of 1
I l
Northeast Utilities ICAVP DR NO. DR-MP3 08M. l Millstone unit 3 Discrepancy Report Review Group: Sptom DR VALID Potential Operab6lity issue Diecipl6ne: Mechanical Detegn Diecrepency Type: Calculabon O ve.
SystenWProcess: RSs i No
~
NRC Significance level: 3 Date faxed to NU:
Date Published: 11/2097 Discrepancy: FSAR Question Q410.5: Waming Time for RWST Overflow Descrippon: In response to FSAR Question No. 0410.5, NU committed to providing a high RWST level alarm which would alarm 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> before overflow at the maximum RWST fill rate.
There is no basis for this function in the QSS system calculations.
The only calculation which addretses this issue, Calculation 357P, Rev. 0 (page 10), states that the high level alarm would provide a 60-minute waraing prior to RWST overflow. This calculation does not provide any documentation to suppirt this statement. The 60-minute overflow waming time is based on a maximum fill rate of 228 gpm from the CHS and SFC systems.
No reference for this fill rate is provide <8. The 60-minute overflow waming time implies that there is 13,680 gal of RWST volume between the high level alarm snd the overflow level. No data is provided for the elevation of the high level alarm setpoint, the maximum setpoint dnft/ error, or the elevation at which overflow would occur. No reference for tank volume as a function of elevation is provided.
A calculation needs to be performed to determine the waming time for RWST overflow provided by the RWST high level alarm.
Review Valid invalid Needed Date inMietor: Wakeaand, J. F.
O O O 11/11/S7 VT Lead: Nort, Anthony A O O O 11'11/S7 VT Mgt: Schopfer. Don K S O O 11/17'S7 IRC Chmn: Singh. Anand K B D 0 11'17'S7 Date:
INVALID:
Date:
RESOLUTION:
Previously identifled by NU? O vos t91 No Non Discrepent Condstion?O vos 'R1 No Resolution Pending?O ve. + No Resolution Unresolved?O yes @ No Review initiator: (none)
VT Lead: Nort, Anthony A VT Mgr: Schopfer, Don K IRC Chmn: Singh, Anand K Date:
SL Comments:
Printed 11/18/97 lt41:07 PM Page 1 of 1
9 Northeast Utilities ICAVP- DR N2. DR MP3-0697 Millstone Unit 3 Discrepancy Report Review oroup: systern DR vAuD 8' Potential Operatnitty issue D6ecipline: Mechancal Des $n y Discrepency Type: Component Data 4~
SystenVProcess: QSS NRC Signincance level:4 Date faxed to NU:
Date Published: 11/2D97 Diseropency: Inconsistency between FSAR Section 6.2.2.2 and spec 2280.000-968 w/ respect to design pH DWr6Pt60n: According to FSAR Section 6.2.2.2, the minimum pH of the spray from the quench spray headers into the containment structure is 4.4. According to design specification 2280.000-968 Revision 10, page 8-15, the quench spray nozzles are designed for a pH range of 4.7 - 10.5.
Review w al6d :nvalid Needed Date inatiator: Feingo6d. D. J- 0 0 0 11/10'S7 11/11/97 VT Lead: Nort, Anthony A Q Q Q VT Mgr: Schopfer. Don K O O O 11'17/S7 BRC Chmn: Singh, Anand K G O O 11/17/S7 Det.:
INVALID:
Date:
REtoLUTION Prov6ously identined by NU7 O Yes t*) No Non Discrepent Condition?C Yes @ No Resolution Pending70 ies + No ResolutionUnresolved?O Yes @ No Review Acceptable Not Acceptable Needed Date VT Lead: Nert, Anthony A VT Mgt: S:%pfer, Don K IRC Chmn: Singh, Anand K Date:
SL Comments:
l 1
l l
l Printed 11/18,97 2:43 31 PM Page 1 ef 1
Northeast Utilities ICAVP DR N3. DR-MP3-0600 Millstone Unit 3 Discrepancy Report Review Group: Accadert Mit$stion DR VALID Potential operatsty issue Diecipline: Mechancel Desqq actwy Typet LW W O Ya (5) No
~
System / Process: N/A NRC segrecance W: 4 Date Faxed to NU:
Date Published: 11/2097 D6screpency: Time Critical Activities Not identified in EOPs or Operator Training Manuals Desert mon: In the response to a Feedwater System Pipe Break, the analysis described in FSAR Section 15.2.8 for a Feedwater System Pipe Break assumes that 30 minutes (Table 6.2 59) is required to isolate the auxiliary feedwater system from the faulted steam generator.
The ICAVP reviewed EOP 35 E 0, EOP 35 E-1, EOP 35 E-2, and the lesson plans used to train the operators for various accident scenarios.
The accident analysis, FSAR Section 15.2.8.2 (page 15.2-19),
assumes that the auxiliary feedwater system is isolated from the faulted steam generator within 30 minutes However, there is no mention of a time requirement in the above procedures or in the training materials. Therefore the operator is not cognizant of a need to isolate the faulted steam generator within a 30-minute period, Review Valid Invai6d Needed Date initiator: Kane.T.J. 8 O O 1 0'30/S7 VT Lead: Rabela, Raj D 8 O O 10/3 /97 VT Mgt: Schopfer. Don K G O O 11'5'87 IRC Chmn: Singh, Anand K O O O 11'17/S7 Date:
INVALID:
Date:
RESOLUTION.
Previously identified by NU? O Yes ei No Non Discrepant Condition?Q Yes is) No b Resolut6on Pending?O Yes '5) No Resolution Unres Wyed?O Yes <s) No Review Acceptable Not Acceptable Needed Date VI Lead: Raheia, Raj D VT Mgr: schopfer Don K IRC Chmn: Singh, Anand K Date:
sL Conwnents:
Printed 11/18/97 2:4452 PM Page 1 of 1
M Northeast Utilities ICAVP DR N2. DR-MP3 0413 Millstone Unit 3 Discrepancy Report Rev6ew Group: System DR VAUD
- M** "
Potential OperabiHty lasue Diecipline: N
- O vos Diecrepency Type: Componert Data
$) No Systen#rocess: OsS NRC Significence levoi: 3 Dete faxed to NU:
Dete Published: 11/2G97 Discrepency: inconsistency between FSAR Sec 6.1.1,1, Tbl 6.1-1, &
component design w/ respect to yeild strength
Description:
FSAR Section 6.1,1.1 states that cold-worked stainless steels exhibiting a yleid strength in excess of 90,000 psi shall not be used. The following are inconsistencies with this requirement:
A. All stainless steel material in the quench spray system has a yield strength less than 90,000 psi, except as follows:
- 1. SA 564 TP 630 Cond H-1150 used for the shaft on valves 30SS*MOV34A,B,30SSV2,6, and 30SS*V945,946.
- 3. SA-193 GR B7 used for the bolts on the RWST have a yield strength less than 90,000 psi only if bolts the are 4 to 7 inches in diameter. The sizes of the RWST bolts are not shown on the RWST drawings. Therefore, the yield strength for this materialis indeterminant. Note that FSAR Table 6.1 1 specifies tank bolting to be SA 193 GRB7.
- 4. SA 193 GR B7 and GR B16 used for bolts on valves 30SS*AOV27/28 have a yield strength '.ess than 90,000 psi only if the bolts are 4 to 7 inches in diameter.
Drawing 2472.110-185-349 shows that the bolts on these valves are SST 18-8. Conflicting data is provided co the valve bolting material.
- 5. A-276 TP 316 Cond B used for the stem on valves 30 SS*V50-58, 940, 941, 948-951, 968-975, 933 936,957 959,964-967,086,987, 42,43.
- 6. A-276 TP 316 Cond B used for the pump shaft on pumps 30SS*P3A,B.
- 7. SA-276 TP 304 Cond A and SA 296 CF8 tsed for th? ctPrpar arm, clapper arm shaft, disc nut pin, and disc nut washer on valves 3QSS*V4,6 and 3OSS*V976-979, and for the impeller in pumps 3QSS*P3A,B. These materials are not listed in ASME Section 11, Part A, eventhough the material designations can be traced back to ASTM specifications A 276, A-743, A 743M A-744, and A.7dakA Pi1nted 11/1&S7 245 28 PM Page 1 of 3
i l
ICAVP- DR No. DR-MP3-0613 Northeast U'ilities Millstone Unit 3 Diccrepancy Report Material references are:
Drawings:
2472.110-185-349 Revision C 2282.050-153-039 Revision B 2362.200164-040 Revision A 2282.150-154-021 Revision G 2282.150154-043 Revision G 2214.602 040-013 Revision C 2362.200164-018 Revision B 2214.601-023-001 Revision R 2282.050-153-036 Revision A Specifications:
2472.110-185 Revision 1 SP ME-784 Revision 2 2362.200164 Revision 1 B. All stainless steel material used in the containment recirculaiton spray system has a yield strength less than 90,000 psi, except as follows:
- 2. SA-564 TP 830, no Condition specified, used for the bonnet nuts and studs on valves 3RSS*MOV38A B.
- 3. A-705 Grade 630, no Condition specified, used for the stem on valves 3RSS*MOV38A,B, 3RSS*MOV8837A,B, and 3RSS*MOV8838A,B.
- 4. A-276 TP 316 Cond B used for the stem on valves 3RSS*V32-33, 58-59, 62-63, 66-67, 913, 916, 920-931, 946-053,988-995, 958-959, 961 962, 64-65,68-69, 912,970-971,973,S77 978,934, 967-968,980-981,911, and 3RSS-V25,941944, 935-938, 918-919, 960, 934, 039, 963-966, 969, 976,979,983.
- 5. A 276 TP 316 Cond B used for the packing gland and packing gland bolts on valves 3RSS*V32-33, 58-59,62-63, 66-67, 913,916,520-931, 946-953, 988-995,958-959,961-962.
- 6. SA 276 TP 304 Cond A. SA-276 TP 316 Cond B, and SA 296 CF8 used for the clapper arm. clapper arm shaft, disc nut pin and disc nut washer on valves 3RSS*V3,6,9,12. These materials are not listed in ASME Section 11, Part A, eventhough the material designations can be traced back to ASTM specifications A 276, A 743, A 743M, A-744, and A-744M.
Printed 11/18f97 2A5:30 PM Page 2 of 3
^
Northeast Utilities ICAVP DR NO. DR-MP3 0613 Ministone unit 3 Discrepancy Report 3RSS*EJ1A,B,C,D and 3RSS*EJ2A,B,C,D. This material is not listed in ASME Section 11, Part A.
Material references are:
z Drawings:
2332.310-669-026 Revision H 2282.150154-019 Revision G 2332.910-669-027 Revision J 2282.150 154-021 Revision G 2282.050-676-103 Revision 1 2282.150-154-022 Revision K 2282.050-676-110 Revisbn E 2282.150-154 026 Revision G 2282.050-153-036 Revision A 2282.150-154-042 Revision G 2282.050153-039 Revislan B 2282.150154-CM3 Revision G 2282.050153-042 Revision A 2282.150-154-044 Revision G 2282.150-154-018 Revision G 2282.150-154-046 Revision
, G Specifications:
SP MP.-784 Revision 2 Revlow Valid invalid Needed Date initiator: Femgold. D. J. O O O 11'50S7 VT Lead: Ner( Anthony A G O O 11/11/S7 VT Mgr: Schopfer, Don K B O O 15'17/97 IRC Chmn: Singh, Anand K Q O O Sil27'S7 Date:
INVALID:
Date:
RESOLUTION:
Previously identtAed by I U? O Yes :O' No Non Discrepent Conddion?Q Yes it) No Resolution Pending?O Ye. + No Resolution Unresolved?O Yes M No Review Acceptable Not Acceptable Needed Date VT Leed: Nort Anthony A VT Mgt: Schopfer, Don K rdC Chmn: Singh, Anand K Date:
SL Comments:
7 teed 11/1&97 245:33 PM Page 3 of 3
l l
. Northeat.t Utilities ICAVP DR N2. Il-MP3-0621 1 Millstone Unit 3 Discrepancy Report Review Group: system DR VALIO Potential Operatsty issue Discipsine: Mechancal Desgn Discrepency Type: Component Data O va
- 5) No SysteriWProcess: Rss NRC Sigencance level: 4 Date faxed to NU:
Date Published: 11/2oS7 piecrepancy: Specs 2214.803-020 & 2214.802-044 are not in agreement with FSAR Se tion 6.5.2.1 w/ respect to pH.
Deecription: FSAR Section 6.5.2.1 states that the containment recirculation spray system is designed to contain a solution with a pH of range of 7.0 to 10.5. However, the following are designed for a pH range inconsistent with the range specified in the FSAR:
- 1. Containtnent recirculation cooleis design speci:lcation 2214.803-020 through Addendum 5, page 1-4, states that the solution on the shell side will have a pH range of 8 to 10.5.
- 2. Containment recirculation spray pump design specification 2214.802-044 through Addendum 6, page 15 states that the pumped solution svill have a pli range of 5 to 7.5.
Review Valid invalid Needed Date initiator: Feingold, D. J. 8 O O 11S7 VT Lead: Nort, Anthony A O O O 11'4S7 VT Mgt; schopfer, Don K Q O O 11'10S7 IRC Civnn: singh, Anand K O O O ' '17/S7 Date:
INVALID:
Date:
RESOLUTION:
Previously identified by NU7 O yes #1 No Non niscrepant ConditiontO yes (#) No Resolution Pending?O yes + No Resosuiionunresolved70 ve. C*)No Review g Acceptable Not Acceptable Needed Date VT Lead: Nort, Anthony A VT Mgt: Schopfer, Don K IRC Chmn: segh, Anand K Date:
sL Comments:
Pnnted 11/1897 2A6.00 PM Page 1 of 1
I l
Northeast Utilities ICAVP DR No. DR-MP3-0624
)
l Millstone Unit 3- Discrepancy Report Review Group: Programmete DR VAUD ,
Potential Operabinty issue piecipune: otne' D6screpency Type: Corrective Acton O vee g~
SystenVProcess: N/A NRC Signincance levet: 3 Date faxed to NU:
Date Published: 11/2097 06.crepency: Storage of Reference Material as QA Records DeectlPilon: Condition Report (CR) M3-971649 deals with faxes and telecons referenced in maintenance procedure MP 3740DA which were not in the possession of document control, but were found in a depar. ment file. In addition, the CR stated that reference material from vendors and outside sources is used without having Engineering evaluation.
No corrective action was taken for this CR since procedure DC 2 titled " Developing and Revising Millstona Procedures and Forms" requires that "any uncontrolled or procedure specific references, such as vendor letters, memes, etc. are to be maintained in department files. The Department that develops and maintains a procedure is responsible for ensuring the availability of all references". In addition the evaluator stated that all referenced material is subject to review by the Engineer w".en reviewing a procedure.
- 1. Not maintaining faxes and te!econs from vendors or other outside sources, which contain technicalinformation used as inputs to procedures, as QA records violates sections 10 and 5.1.4 of ANSI N45.2.11 1974 which is a commitment of the NU Quality Assurance Program Topical Report. Section 10 states in part:" Design documentation and records which provide evidence that the design and review process was performed in accordance with the requirements of this standard shall be collected, stored and maintained in accordance with the requirements of ANSI N45.2.9. The documentation shallinclude not only the final design documents such as drawings and specifications, and revisions thereto but also records of the important steps including sources of design input ...". Section 5.1.4 states in part: "Where it is necessary to initially transmit design Information orally or by other informal means, the transmi'tal shall be confirmed promptly by a controlled document."
- 2. It is not clear that the engineering review of information received from vendors and other outside sources complies with the Design Control Manual (DCM). Section 1.3 of the DCM, revision 5, Chapter 8, assigns responsibility for Unit specific Engineering vendor interfaces to the Design Engineering Manager. Section 4.0 of Chapter 8 requires the Engineer to review or coordinate the review of vendor design documents and other eng!neering deliverables. Table 1 of Procedure DC 3 titled
" Verification and Validation of Procedures and Forms" requires an engineering review of procedures (including inputs) only when determined necessary by the Responsible Individual, the Department Head or the Department Procedure Coordinator.
Printed 11M8,97 2:47:37 PM '
age 1 of 2
ICAVP DR No. DR-MP3 0624 Northeast Utilities Millstone Unit 3 Discrepancy Report i
technical inputs, received from vendors or outside sources, be l reviewed by Engineering. Thus, it is not evident that l Engineering reviews all vendor techr.ical information affecting l proceauret.
Not storing the above described reference material as QA ,
reccMs and the ambiguity concerr'ing the engineering review of (
this material constitutes a discrapancy.
Review Val 6d inval6d Needed Date inMietor: Shepperd, R. P. O O O 1/7/S7 VT Leed: Ryan. Thomme J @ Q Q 11/7/97 VT Mge: Schopfer, Don K G O O i/10S7 IRC Chmn: Singh, Anand K Q Q Q 11/17/97 Date:
INVALID:
Date:
RESOLUTION:
Previously identmed by NU? O vee t#) No Non Discrepent CondM6on?O vee f9) No ResolutionPending?O vee 4) No ResoiviionunresoivedtO vee @ No l Review ACC*Ptable Not Accogdeble Needed Date g,
VT Lead: Ryan, Thomme J VT Mgt: Schop'er, Don K IRC Chmn: Singh, Anand K Date:
SL Comments:
Prtnted 11/18,97 2.47:42 PM Page 2 of 2
ICAVP DR No. DR-MP3-0631 Ncrtheast Utilities Millstone Unit 3 Discrepancy Report Review Group: System DR VALID Rh EW: S#em %n Potential Operobilsty issue ^
Disc 6pline: Mechanical Design Discrepency Type: Component Data Om (e--) No Systemfrocess: QSs NRC 56gnificance level: 3 Date faxed to NU:
Date Published: 11/2orJ7 Descrepancy: Inconsistency between FSAR Table 6.1 1 and material design specs and drwgs w/ respect to materials D**cription: ltem i FSAR Table 6.1 1 specif:es bolting and studs for engineered safety system components to be either material specificetions SA 193 Grade B6 or Grade 87, or SA-452 Grade 660. Bolting material given in FSAR Table 6.1 1 does not specify whether it applies to all bolting or just pressure boundary botting.
Therefore, the material specifications in the FSAR are assumed to apply to all botting.
The following components use bolting from material other than SA-193 Grades B6 or B7, or SA-452 Grade 660:
l 3QSS*AOV27 (has SA 193 B7 bolting, also) 3QSS*AOV28 (has SA-193 B7 bolting, also) 30SS*V4, V8 3QSG'V976-V979 3OSS*V1, V5 3QSS*V50-58 3QSS*V940, V941 3QSS*V948-V951 3OSS*V968-V975 3QSS*V933-V936 3QSS*V957 V959 3QSS*V964 V967 3QSS*V986, V987 30SS*V42, V43 Material references are:
Drawing 2362.200164 040 Revision i Drawing 2362.200164-018 Revision E Drawing 2282.050-153-036 Revision A Drawing 2282.050-153-039 Revision B Drawing 2282.150154-021 Revision G Drawing 2282.150154-043 Revision G Drawing 2472.110-185-349 Revision i Specification 2472.110-185 Revision 1 Note that some of the pressure boundary bolting shown on the design drawings and specificadons is SA-453 Grade 660, so there might be a typographical error associated with this botting.
Piinrod 11/18S7 2:48.19 PM Page 1 of 3 i
' DR N3, DR-MP3-0631 Northeast Ut"' ties ICAVP Millstone Unit 3 Discrepancy Report item 2 FSAR Tables 6.1 1 shows the stainless steel valves and components in the engineered safety features systems to have nuts made of specifications SA 194 Grade 6, SA-194 Grade 2H or SA-453 Grade 660 mateiial. The nut material given in FSAR Table 6.1 1 does not specify whether it applies to all nuts or just pressure boundary nuts. Therefore, the material specifications given in the FSAR are assumed to all nut material.
Tiie following components have nuts from material different than that specified in FSAR Table 6.1-1:
3OSS*AOV27 3OSS*AOV28 3QSS*V4, V8 .
3OSS*V976 V979 3QSS*V1, V5 3OSS*V50-58 3QSS*V940, %#1 3OSS*V948-V9:,1 3OSS*V968-V975 3QSS*V933-V936 3OSS*V957 V959 3OSS*V964 V967 30SS*V986, V987 3QSS*V42, "43 Material references cre:
Drawing 2362.200-164-040 Revision i Drawing 2362.200-164-018 Revision E Drawing 2282.050153-036 Revision A Drawing 2282.050-153-039 Revision B Drawing 2282.150154-021 Revision G Drawing 2282.150-154 043 Revision G Drawing 2472.110-185-349 Revision 1 Specification 2472.110-185 Revision 1 Review Valid I,ivalid Needed Date inteletor: Feingold. D. J. O O O iii12s7 VT Lead: Nort. Anthony A O O 1 /12/97 O
VT Mgt: Schopfer. Don K O O O 11'17/97 IRC Chmn: Singh. Anand K g Q O 11/17S 7 Date:
INVALID:
1 Date:
l RESOLUTION Previously identifled by NU7 C) Yes @ No Non Discrepant Condition?C) Yes #1 No l
Resolut6on Pending70 Yes @ No Resolution Unresolved 70 Yes @ No Printed 11/18/97 2:48 23 PM Page 2 of 3 t
l
- DR N3. DR MP3-0631 Northeest Utilities ICAVP Millstone Unit 3 Discrepancy Report Acc e .u. uw Acc., w. Revi.ew u o,,,
w w m (non,3 VT Lead: Neet, Anthony A O O VT Mge: Schopfw, con K O O 1RC Chmn: Sangh, Anend K O G O O 8 om.:
SL Comments:
l l
Printed 11/18,97 2 40 25 PM ~ Page 3 of 3-
ICAVP DR N . DR-MP3-0641 Northeast Utilities Millstone Unit 3 Discrepancy Report Rev6ew Group: Pry...nehe DR VALID 3 I
- Potential Operability issue Disc 6pline: Mechanical Desagn O ve.
GE - ay Type: Test Requirements 4~
SystemProcess: N/A NRC Significance level: 4 Date faxed to NU:
Dete Published: 11/20/97 D6ecrepancy: Use of Magnaflux SKC-NF Cleaner / Remover Descript6on: Change Notice NU-LP-1 1 dated 11/10/93 to Procedure NU-LP.
1, Revision 11, titled ' Liquid Penetrant Procedure, Color Contrast Solvent Removable" was reviewed as part of the Tier 3 review of past procedure changes for technical adequacy and compliance with the licensing bases. This change allowed the use of Magnaflux SKC-NF cleaner / remover provided the examir.ation was followed by a flush with deionized water and a wiping by a fresh cloth or absorbent paper.
SKC NF is a general purpose remover which is chlorinated and not recommended by Magnaflux for nuclear applications.
Paragraph T-644 of ASME Section V requires that NU obtain certification by batch number of the centaminant content of all liquid penetrant materials used on nickel based alloys, austenitic, stainless steels, and titanium. In addition, T-644 requires that penetrant materials used to examint nickel based alloys have a sulfur content less than or equal to 1% (by weight) of a residue, when the residue left by the specified test exceeds 0.005 g. T-644 also requires that penetrant materials used to examitte austenitic stainless steels o, titanium have a chlorine content less than or equal to 1% (by weight) of a residue, when the residue left by the specified test exceeds 0.005 g.
It is not clear from the procedure change whether tne SKC-NF material was analyzed for the above constituents and subsequent controls established depending on the results of the tests.
Review Valid invalid Needed Date inMlator: sheppard, R. P. O O O ' 15 S 7 O 1 /SS7 VT Lead: Ryan, Thomas J G O VT Mgt: schopter, Don K B O O 11/10/S7 11/17,97 IRC Cimm: singh, Anand K @ Q O Date:
INVALID:
Date:
RESOLtJTION Prev 6ously identifled by NU? O Yes @ No Non6,tas , pant Conddion?O Yes ? No Resolution Pending?O vos a No Resolution unredved?O Yes M No Review Acceptable Not Acceptable Needed Date VT Lead: Ryan, Thomas J Printed 11/1&97 2:5o:00 PM Page 1 of 2
ICAVP DR No. DR MP3 0641 Northeast Utilities Millstona unit 3 Discrepancy Report i
. . . m r. s . . . .
VT Mgt: Schopfw, Don K O O O IRC Chmn: Singh, Anand K O- 0 0 '
O O O
,,. l SL Comments, l
Printed tin &S7 2.50:04 PM Page 2of 2
4 ICAVP DR N2. DR-MP3 0643 Northeast Utilities Discrepancy Report Millstone Unit 3 -
Rev6ew oroup: System DR VALID N'* Potential Operability issue Diecipl6ne: Mechenecal Des
- O Yes Discrepency type: Component Data
'5) No
~
System @rocess: Rss NRC Sign 4Acance level: 4 Date faxed to NU:
Date Published: 11/20/97 D6ecropency: Inconsistency betwn FSAR Sec 6.2.4.1.1.2 & spec SP ME 784 for containment isolation viv design press
Description:
According to FSAR Section 6.2.4.1.4. 2, the design pressure of all piping and components within the isolation boundaries
, afforded by the containment isolation system is equal to, or greater than, the design pressure of the reactor containment.
According to FSAR Section 6.2.1.1.2, the containment design pressure is a psia to 45 psig. According to FSAR Section 6.2.1.1.3.7, the containment liner design temperature is 280 degrees Fahrenheit.
The valve data sheet for valves 3RSS*MOV23A,B,C,D in design specification SP-ME 784 Revision 2 show the design pressure and temperature of valves to be 40 psig and 280 degrees Fahrenheit. However, the same specification identifies the valve as being 150 # class with a valve body constructed to material specification SA 351 CF8M. According to ANSI B16.34, Ratings For Group 2 2,1 Materials, the rating for these valves is 205 psig at 300 degrees Fahrenheit.
Valves 3RSS'MOV23A,B,C,D are capable of withstanding an operating pressure greater than the containment design pressure based on the valve ANSI B16.34 design class, but the specified design pressure is less than the containment design pressure.
Review Valid invalid Needed Date initiator: Feingold. D. J. O O O i ri2/97 VT Lead: Neri, Anthony A O O O iiri2/97 VT Mgr: schopfer, Don K O O O 11/17/97 IRC Chmn: singh, Anand K O O O 1 '17/S7 Date:
INVALID:
Date:
RESOLUTION Previously Idem. *ed by NU? O Yes ?! No Non Discrepent Condstlot ?Q Yes ? No Resolution Pending?O Ye. s) No Resolution Unresolved?O Yes (5) No Review Acceptable Not Acceptable Needed Date VT Lead: Nert. Anthony A VT Mgt: schopfer, Don K j P8tC Chan: singh, Anand K Date: l Printed 11/1tV97 2.5o1. PM Page 1 of 2 )
l l
l
Northeast Utilities ICAVP DR Ns. DR-MP3-0643
-:Miiistone Unit 3 Discrepancy Report SL Comments:
e Printed 11/1&S7 2.5123 PM Page 2 of 2 l
ICAVP DR No. DR-MP3-0649 Northerst Utilities Millstone Unit 3 Discrepancy Report Revlow Oroup: System DR VALID Review Element: System Gesign pgg g,,
PE ', - . Mechancel Desgn D6screpancy Type: Licensing Documett O Ya t#) No
' ~ '
SystemP ocess: QSS NRC Sign 4Acance level: 4 Date faxed to NU:
Date Published: 11/20/97 Dwncy: Length of Pipirig Between QSS Containment isolation Valve and Containment Wall D**cription: FSAR Sec. 6.2.4.1.4, item 4 and FSAR Table 6.2-66 require that there be no more than 10 feet of piping between QSS containment isolation valves and the contain, ment wall.
According to Piping Drawing 12179-EP-79G, Rev. 9 there is 11'.
O' of piping between QSS coritainment isolation valve 3OSS*MOV348 and the coruainment penetratioi1 (the containment penetration pratrudes 6 or 10 inches from the outside of the containment wall).
Rev6ew Valid invalid Needed Date instietor: Wakeland, J. F. O O O 11'11/97 VT Lead: Nort Anthony A B O O 11'SS7 VT Mgr: Schopfer, Don K -
0 11'17/S7 IHC Chmn: Singh, Anand K -
O O 11'17/S7 Date:
INVALID:
F Date:
RESOLUTION:
Previously identified by NU? O Yes '41 No Non Discrepent Condet6on?Q Yes #1 No Resolution Pending?O vos *> No Resolut6on Unresolved?O yes 6) No Review Acceptable Not Acceptable Needed Date VT Lead: Nort, Anthony A VT Mgr: Schopfer, Don K IRC Chmn: Singh, Anand K Date:
SL Comments:
Printed 11/1&97 2.51:42 PM Page1 of 1
Northeast Utilities ICAVP DR Nr. DR MP3-0660 Millstone Unit 3 Discrepancy Report Revtew oroup: Contguraten DR VAUD Re*w Elenwnu systern Desgn Potential Operabmty issue
. Diecipline: Piping Desag" D6screpency Type: Drawmg Om fe) No
~
SystemfProcess: DGX NRC Sign 6ficence level: 4 Date faxed to NU:
Date PutAlshed: 11/2097 D6scr*Pency: Upper Tier to Lower Tier Drawirig Review for DGX Description
- The following drawing discrepancies were found in reviewing the P&lD with the as built Isometrics for the Emergency Diesel Generators (DGX):
- 1. On isometric CP 360258 Sht 4 Rev 6 line 3 EG A500-61-3 should be 3-EGA 500-64 3,line between valve *W95 and '
downstream reducer should be labeled 3-F.GA-500-913, and line between air receiver 3EGA*TK1 A and first downstream flange should be labeled 3-EGA 500-90-3 per P&lD EM 116B Rev 28 (1-6).
- 2. On isometric CP 360259 Sht 4 Rev 8 the line between air receiver 3EGA*TK2A and fi% downstream reducer should be labeled 3 EGA-500-88 3 and the line between valve 3EGA*V994 and first downstream reducer should be labeled 3-EGA 500-89-3 per P&lD EM 116B Rev 28 (1-8, H-9).
- 3. On support drawing BZ-60R-63-1 Sht 1 the supported line called out as 3-EGA-375-25-3 should be 3-EGA-375-17-3 per P&lD EM-116B Rev 28 and CP 360259 Sht 4 Rev 8.
- 4. On isometric CP 360257 Sht3 Rev 4 line 3-EGA-250-62-3 should be 3-EGA 500-62 3 in the line list box. Line 3-EGA 375-19-3 should also be listed.
- 5. On isometric CP 360260 Sht 4 Rev 5 the line between air receiver 3EGA*TKiB and first downstream reducer should be labeled 3-EGA 500-86-3 and line between valve 3EGA*V982 and first downstream reducer should be labeled 3-EGA-500-87 3 per P&lD EM 11rO Rev 9 (1-6). Also the drafting symbol for the excess flow check valve 3EGA*V984 is shown as a standard check valve (see CP-360258 Sht 4 *V986)
- 6. Flex hoses 3EGA* Hose 1B/2B are not listed in GRITS.
- 7. Valves 3EGA*V9, 3EGA*V995,EGA*V5, 3EGA*V13, 3EGA*V4, 3EGA*V11,3EGA*V16, 3EGA*RV24 A1/A2, 3EGA*V30, 3EGA*V37, 3EGA*V38, 3EGA*V31, 3EGA*V981, 3EGA"V982,3EGA*V42,3EGA'V35,3EGA*RV24B1/B2 and
" 3EGA*V994 are not listed in GRITS and hence.no vendor drawings could be identified.
- 8. On isometric Cl EGD-501 Sht3 Rev 2 safety related equipment 3EGD*EJ2A,*EJ2B,*EJ2C, and EJ2D per P&lD EM-116E Rev 2 should be designated with a (*).
Printed 11/18,97 2.52-34 PM hEof3
Northeast Utilities ICAVP DR N3. DR-MP3 0660 Millstone UnN 3 Discrepancy Report 3EGD*V2, 'V1, and diesel ge ierator 3EGS*EG-A/B per P&lD EM 116E Rev 2 (17, F 7) sho@i be designated with a (*).
- 10. On isometric CP 360503 sht 3 Ric. G Uie embedment s:'s dd not be shown on line 3-EGF 002 7 3 until after connection to line 3-EGF-002-43-3 per P&lD EM-117A Rev10 (F 3).
- 11. On isometric CP 360503 Sht3 Rev8 line 3-EGF-001 15-3 should not show embedment per P&lD EM117A Rev10 (G-2) and the isometric should be continued through line 3-EGF 002-43-3 as shown on CP360503 Sht 1 Rev 8,
- 12. On isometric CP-360530 Sht 10 Rev 1 restricting orifice 3EGF*RO39A is shown but not labeled as per P&lD EM 117A i Sv 10 (L-3). Also not listed in GRITS.
- 13. On isometric CP 360533 Sht 19 Rev 8 restricting orifice 3EGF*RO39B should be labeled as per P&lD EM 117A Rev 10 (1.-7). Also not listed in GRITS.
- 14. On isometric CP 360534 Sht3 Rev 7 line labeled 3-EFG-001-213 should be 3-EGF 001213 per P&lD EM 117A Rev 10 (H-8).
- 15. Valves 3 EGO *V986A, 'V989A, *V990A, 'V982A, *V16A,
'V981 A, 'V18A, 'V988A, 'V980A, *V987A, 'V991 A, *V985A,
'V984 A, *V983A, *V17A, *V992A, 'V986B, *V989B, *V990B,
- V982B, 'V16B, 'V981B, *V188, *V988B, 'V980B, 'V987B,
- V991B, *V985B, *V984B, *V983B, 'V17B and *V9928, *V20B are not listed in GRITS and hence no vendor drawings could be idtatified.
- 16. On isometric CP-360264 Sht 4 Rev 5 line between flex hose 3EGS* Hose 4A and the downstream tee should be labeled 3-EGS-375113 per P&lD EM-116A Rev 30,
- 17. On isometric CF-360700 Sht 3 Rev 5 the line between the Diesel A and first coupling on line 3-EGS-375-9-3 should be labeled 3-EGS-500 22 3 per P&lD EM-116A Rev 27.
Review Valid invalid Needed Date initiator: Road. J. W. G O O 11'1157 VT Lead: Neft. Anthony A Q Q Q 11/11 S 7 VT Mgr: schopfer, Don K G O O $1'17/97 0 IRC Chmn: Singh. Anand K Q Q Q 11/17/97 Date:
INVAUD:
Date:
REs0LUTION:
Prevk>uely identined by NU? Q Yes 9) No Non Discrepant Condd6on?Q Yes * > No Pesolution Pendmg?O Ye. + No Resolution Unresolved?O Yes @ No Review Printed 11/18S7 2.52;38 PM Page 2 of 3
Northert Utilities ICAVP DR N3. DR-MP3-0650 Millstone Unit 3 Discrepancy Report Acce.m woi acc e . m u 4.d De VT Le d: Neri, Antnany A O O O vi m or: senope.r, con x 0 0 0 RC Chmn: Singh, Anand K Date:
SL Conn.nts:
(
Prtnted "*%7 2.52:40 PM Page 3 of 3 l
l
ICAVP DR N3. DR-MP3-0652 Northeast Utilities Millstone Unit 3 Discrepancy Report Review GrOJp: Confgurata DR VALID Potenhal OperatAllty lasue D6scipline: Peng Dugn O va D6ecrepency Type: Instanatm implementate g" SystenVProcess: DGX NRC Significance level: 4 Date faxed to NU:
Date PeMt ittXV97 Descrepency: Walkdown Discrepancy for the DGX h Ducripuon: The following discrepancy items were found during the walkdown of the piping and mechanical equipment of the Emergency Diesel Generators (DGX):
- 1. Line 3-EGA 750-13 3 is not insulated as required by spec SP-ME-691.
- 2. Line 3-EGA 375-30 3 is not insulated as required by spec SP-ME-691.
- 3. Flow element 3EGF*FE31 A mfg. data (bore) is illegible due to label being painted.
- 4. Restricting onfice 3EGF*RO20C mfg. data (bore) is illegible i due to the label being painted.
- 5. Restricting orifice 3EGF*RO20A mfg. data (bore) is illegible.
- 6. Flex hoses 3EGS* Hose 3A and 3EGS* Hose 4A NU labels should be interchanged to be consistent with isometric CP-360264 Sht 4 Rev 5 and P&lD EM-116A Rev 27.
- 7. Flex hose 3EGS* Hose 6A has no NU label.
- 8. Pipe support CP-360530-H005 st.own on drawing BZ-60R-40 Rev 2 has two additional iin x iln bars welded to the support to provide lateral restraint to the pipe that are not shown on the drawing. Unincorporated E&DCR T-J 06448 (copy obtained from NU's drawing vault) has incorrect subsequent pages attached to its cover page and hence could not be used to determine if the additional bars are correct.
The following material condition items were found during the walkdown of the piping and mechanical equipment of the Emergency Diesel Generators (DGX);these items are not considered configuration management issues:
- 1. Valves 3EGF*V996 and 3EGF*V997 packing shows leakage.
- 2. Valves 3EGF'V25, *V26 and*V22 stems and bodies are rusted.
- 3. Valves 3EGF'V11 and 'V14 packing show leakage. Also the
- mfg. label are not legible and not all of the data could be verified.
' 4. The threaded connection on line 3-EGF 150-48-3 to the diesel
'~ # '
Pnnted 11/18.97 2:53.15 PM Page 1 of 2
~
Northeast Utilities ICAVP DR No. DR MP3-0662-Millstone Unit 3 Discrepancy Report
- 5. The threaded connection on line 3-EGF 00159-3 to the diesel upstream of valve 3-EGFN969 shows leakage.
- 6. The threaded connection on hne 3-EGF-150-49-3 to the diesel downstream of valve 3-EGFN989 shows leakage.
- 7. The threaded connection on line 3-EGF 150-45-3 to the diesel upstream of union shows leakage.
- 8. Tubing connections at valve 3EGFN9L 1, at tee upstream of instrument maniiGid for LS-41B, and at instrument manifold for LS-408 show leakage.
- 9. Threaded connections on strainers to valves EGON9858, N984B and N983B show it,dkage.
Review Valid Invalid Needed Date initietor: Reed, J. W. O O O 115/S7 VT Lead: Nort. Anthony A O O O 11'11'S7 VT Mgt: Schopfer. Don K O O O 15' 7'S7 O O 117'S7 IRC Chmn: Singh, Anand K O Date:
INVALID:
Date:
RESOLUTION.
Previous y idenOAed by NU7 O Yes i91 No Non Discrepent Condstion?Q Yes (9) No Resolution Pending?O Yes 'M No Resoludon Unresolved 70 Y.. '#) No Review Acceptable Not Acceptable Needed Date VT Lead: Neft, Anthony A VT Mgt: Schopfer. Don K IRC Chmn: *, ingh, Anand K Date:
SL Conwnents:
Pnnted 11/1tL97 2.53.19 PM Page 2 of 2
4 Northeast Utilities ICAVP DR No. DR-MP3 0610 Millstone Unit 3 - Discrepancy Report Review Group: System DR INVALID s
Potential Operability issue Diecipline: Electncal Design Discrepency Type: Ucensing Document Om SysterrvProcess: HVX g
NRC Sign 4Acance level: 3 Date faxed to NU:
Date Published: 11/20/97 Discrepancy: Discrepancy between FSAR heat tracing reqmt & design for Rad Monitors 3HVR*RE10A & B sample lines Deecription:
Background:
Page 8.3-29 of the FSAR, Section 8.3.1.1.4 Design Criteria states:
Heat Tracing - The majority of safety related lines and valves are located in heated areas and are not subject to freezing. All safety related lines or valves which are subject to freezing or boron precipitation are electrically heat traced and insulated.
Each such line is electrically heat traced by two circuits, each of 100 percent capacity, with one designated as the normal circuit and the other as standby. On any safety related line that is heat traced, each normal and standby circuit is connected through isolation transformers to the Class 1E Civision "A" or Division "B" bus respectively. In the event of a loss of normal ac power, each emergency bus is carried by its own emergency generator, thereby providing a separate power source to 3ach heat tracing circuit on each safety related line.
P&lD EM 148A, Rev. 24 - lines 3-HVR 750-40-4,3-HVR 750 4,3-HVR 750-14, and 3-HVR-006-30-4 are shown heat traced.
These are the sample lines for radiation monitor 3HVR*RE10A/B, a Reg Guide 1.97 type E2 variable.
Per Thermon Manufacturing Company drawing 80-116, Rev. O, line 3-HVR-750-14, which is the tap to radiation monitor 3HVR*RE108, is heat traced by only one circuit. There is no apparent backup to the heat tracing for the line.
Per Thermon Manufacturing Company ,'rawing 80-116, Rev. O, line 3-HVR 750-2-4, which is the tap to radiation monitor 3HVR*RE10A, is heat traced by only one circuit. There is no apparent backup to the heat tracing for the line.
Both of these heat trace circuits are powered from 3NHS-MCCSB2, a bus that is unpowered in the event of a loss of normal ac, power.
Discussion:
The lines in question, which are taps from the stack to the radiation monitors, run outside along the roof and are subject to freezing if the heat tracing fails. If the lines plug thet, the rndintinn mnniinre whir'h nrn rannired int nff.ci?n dnen Pnnted 11/16S7 4 26.18 PM Page 1 of 3 l l
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Northeast Utilities ICAVP DR No. DR-MP3-0610 Ministone Unit 3 Discrepancy Report monitoring, will not function property.
For the design shown on P&lD EM 148A, the lines do not appear to have redundant electric heat tracing and the source of power, for both circuits, is from a non safety related bus. This appears to be in conflict with the FSAR.
Drawings Reviewed:
EE-9EJ Rev,12 Wir Diag MCCSB2 Sh.1 Rod Cortrol Area EE-63AE Pev.2 Wiring Diag Heat Tr Sys Panel 3 HTS-PNLA4 EM 148A Rev.24 Piping & Instrumentation Diagmm Reactor Plant Ventilation EM-148D Rev.8 Piping & Instrumentation Diagram Reactor Plant Ventilation 80 116 Rev.O Reactor Plant Ventilation System -
Thermon Manufacturing Company "aan comment"sorry but i noticed the lines are class 4 or i non safety related. we may have another issue here, talk to stout. we need to know if these lines are supposed to be safety related. right now the lines are classified as nsr and therefore i heat trace is not in conflict with fsar. by the way,the last digit in line no is safety class.
Review Valid invalid Needed Date inittetor: Womer, l. O O O 117/S7 VT Lead: Nert, Anthony A O O O 11' 7/S7 VT Mgr: schopter, Don K O O O IRC Chmn: srgh, Anand K Q Q O Dete: 11/17/97 INVALID: Lines 3-HVR 750-1-4 and 3-HVR-750-2-4 are not safety related lines, therefore there is no requirement that redundant heat tracing be provided.
For additional information conceming these lines see DR-MP3-0656.
Date:
RESOLUTION:
Previously ident6 fled by Nu? O Yes t@ No Pett Dit:ttpot Ocr.dstion?Q ves (*) No Resolution Pendmg?O yeo (itNo Resolution Unresolved?O vee rf) No Review g Acceptable Not Acceptable Needed Date VT Lead: Nort. Anth:my A VT Mgr: schopier, Don K me r-. m_
' ~
O O O Printed 11/16/97 4 M28 N ~ Page 2 of 3
ICAVP DR No. DR MP3-0610 Northeast Utilities Millstone Unit 3 Discrepancy Report
_ ......... ~-~^
O O O Date:
SL Comment.:
Pnnted 11/18.97 4 26.31 PM Page 3 el 3
Northeast Utilities ICAVP DR No. DR-MP3-0162 Millstone UnM 3 Discrepancy Report Rev6ew Group: Accadert Mtgetson DR RESOLUTION ACCEPTED I O " Potential operataty issue .
Discipline: ONr Diocrepancy Type: Licensing Cocument Om
@ No SysterrvProcess: N/A NRC $6 9 t#cance level: 4 Date faxed to NU:
Date Published: 9/2497 Descrepency: FSAR Inconsistency With Calculation of Record DesertPtion: A review of ifie following documents has concluded that an inconsistency exists between the calculation of record for the locked rotor accident (N Loop and N-1 Loop cases) radiological analyses and the Millstone 3 (MP3) FSAR. The documents reviewed are:
- 1) MP3 FSAR Table 16.3-3 .
- 2) 89-012101RA, Rev,0 - Doses to Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) from a Locked Rotor Accident, N Loop Operation
- 3) 89-012102RA, Rev. 0 Doses to EAB and LPZ from a s
Locked Rotor Accident, N-1 Loop Operation FSAR Table 15.3-3 indicates that for the N-Loop and N-1 Loop cases the activity released to the reactor coolant from failed fuel is 3% of the noble gas and lodine inventory. This is inconsistent with the values in the referenced calculations which are 6% and 4.4% for the N Loop and N-1 Loop cases respectively. The ieferenced calculations were performed specifically to address new failed fuel fractions identified in the MP3 Cycle 3 reload analyses. The MP3 FSAR has not been updated to reflect these values.
l Raiew Valid invalid heeded Data Indiator: Bennett, L A . O O O 9 9/97 VT Lead: Rahoja, Raj D G O O S'1oS7 VT Mgt; schopfer, Don K O O O S/1SS7 BRC Chmn: sogh, Anand K O S':5'S7 O O Date:
INVAUD:
Date: '1f//g7 RESOLUTION: Disposition:
NU has concluded that the issue reported in Discrepancy Report, DR-MP3-0162, does not represent a discrepant condition. Refer f to note 4 on Table 15.3-3. This note explains that while 3% is
? given in the Table, the actual values used in the dose calculations presented in Table 15.0-8 are 6% for N loop and 4.4% for N-1 loop. As a result of the 50.54(f) effort, recommended changes are to be submitted for Table 15.3-3 in FSAR CR 97 MP3-426, but the 3% values were not changed. As !
explained in Note 4 to Table
~
15.3 3. the values in the Table are Printed 11/1&97 4.28.51 PM Page 1 or 2
Northeast Utilities ICAVP DR No. DR4483 0162 Millstone Unit 3 Discrepancy Report those used in detailed calculations petiormed when the calculated radioactivity released was ihree percent of the GAP ,
activity and the steam gi nerator isolation time was 30 minutes.
The doses presented in Table 15.0 8 are based on the 6% and 4.4% GAP activity and a 20 minute isolation time. Proposed changes made to Table 16.3 3 in FSAR CR 97 MP3-426 more clearly identify the note as it pertains to GAP activity and steam generator release times.
A prelimiaary copy of FSAR CR 97 MP3 426 was attached to M3-IRF 00291 (ICAVP response to DR MP3 0016).
Significance Level enteria do not apply here as this is not a discrepent condition.
Conclusion:
NU has concluded that the issue reported in Discrepancy Report.
DR MP3-0162, does not represent a discrepant candition. The dose calcuiation used the correct values for Qap activity released and the stoam Generator isolation time. This is denoted in note 4 of Table 15.3 3. As annotated it. FSAR CR 97 MP3 426, the Table will be revised % more clearly reference Note 4 (which will change to Note 2) and avoid future confusion. Signifiesnce Level criteria do not apply here as this is not a discrepant condition.
Previously identified by Nu? fGI Yes O No Non D6ecrepent Condmon?Q Yes 3) No Resolution Port' .17 0 Yes #) No Resolution Unresolved?O ve. 4) No Review initiator: Bennett. L A.
VT Lead: Rahes,RajDs VT Mgr: Schopfer, Don K IRc Chmn: Singh. Anand K Dele: 11D/97 sL comments: ICAVP agrees that this was a discrepant condition which was identified by NU and corrected in FSARCR 97 MP3 426.
However, we suggest that the revised footnote be annotated by the 3% value ir, TaNe 15.3 3, such that FSAR users will note the j differing analysis cases.
l l
l l
1 i
Pnnled 11/1&'97 426 50 PM Pope 2 of 2
ICAVP DR No. DR MP3 4329 Northeast Utilhies Millstone UnN 3 Discrepancy Report Review Group Prcyemmeno oR RESOLUTION ACCEPTED Rev'ow Element: Change Process , ,
D6ecip46ne: Other O vos D6ecrepency Type: Procedure imp 6ementation ,g g system / Process: N/A NRC 569 n6Acancelevel 4 Date FAKM to NU:
Date Published: 10/1D97 l D*ecropency: Final Safety Analysis Report Change Request (FSARCR) technicalieview documentation Descripuon: The following requirements apply to technite.1 reviews:
A. Section 5.2.15 of ANSI N18.7, which is a comittment in NUQAP Topical Report, requires in part. that measures assure tlist documents are reviewed for adequacy by appropriate personnel.
B. Nuclear Group Procedure (NGP 4.03 Revs. 8 & 9) requires the PRD and additional Reviewers to complete and sign block
- 14. (Refer to Section 6.4, Task 10 of the NGP).
- 1. Contrary to the above, two persons signed in block 14 for FSARCR 97 MP3-30 whose names are different than the designated PRD Reviewer. It is not evident whdher these persons individually or collectively have the required qualifications / authority to sign for the designated PRD Reviewer.
- 2. Contrary to the above, one person whose narae is designated as a PRD Reviewer on FSARCR 97 MP3-68 did not sign. It is not clear that the person that did sign (who was also specified as a PRD Reviewer) has the required qualifications / authorization to sign for the person who did not sign.
Review Volid inval6d Needed Date inattetor Neverro, Mark g Q Q STS97 VT Lead: Rya.,, Thomas J Q O O S2SS7 VT Mgt: Schopfer. Don K O O O $0/5/87 IRC Chmn: Singh, Anand K O O O 100./87 Date:
IlMLlo:
Date: 11/12197 i REsot.UTION: Disposition NU has concluded hat the issue reported in Discrepancy Report, DR MP3-0329, does not represent a discropant condition.
- Section 5.3 of NGP 4.03 requires the manager of Nucleor Licensing (NL) to maintain a listing of the Primary Responsible Discipline (PRD) who is accountable for the accuracy of information in particular sections of the FSAR, and to publish this responsibility listing periodically to appropriate concemed persot,nel, in addition,it is the responsibility of the NL manager to ensure that FSAR changes are processed in accordance with this procedure. Attachment A is a copy of the present PRD Printed 11/1897 4 27.s0 PM Pope 1 of 2
- m. __. __ _ _ _ __ ____
n ICAVP DR No. DR-MP3 0329 Northeast Utilities Mill tone UnN 3 Discrepancy Report Listing distributed by memo NL.97193 dated July 30,1997.
Attachment B is a copy of the PRD Listing distributed by memo NL.97-051 dated March 3,1997. Attachment C is a marked up copy of the PRD Listing maintained by NL showing changes in responsibility which took place from 3/3/97 to 7/30/97. As shown on page 2 of the Attachment C. M. A. Powers and D. M. Aube replaced C. J. Ashton (resigned from NU) and L. Meter (resigned from NU) as the responsible person. Since both of the FSARCR's were initiated about the time these employees were leaving, their names (Ashton & Meter) were placed in block 14 by NL as the PRD per task 8 of NGP. It is agreed that in most cases, the name printed in block 14 of the FSARCR form should agree with the PRD listing, however in all cases, the PRD listing maintained by NL is the controlling document. These signatories of the FSARCR's were qualified / authorized to sign the referenced changes.
Significance Level critoria do not apply here as this is not a discrepant condition.
Conclusion NU has concluded that the issue reported in Discrepancy Report, DR MP3-0329, does not represent a discrepant condition. The designated PRD Reviewer is the person or persons documented on the FSAR Primary Responsible Discipline iist (see attact.ments) maintained by Nuclear Licensing. The two names printed in block 14 were replaced on this list by the names signed in block 14 and the final FSARCR was reviewed by the NL manager to ensure the process was in accordance with procedure NGP 4.03. Significance Level criteria do not apply hers as this is rot a discrepant condition.
Previously idenut,ed by Nu? O Yes 19i No Non Discrepent CondM6on7 97 Yes O No Resolut6on Penvi6ng70 Ye. si No Re.oivison unre.oived70 va.
- No Rev6ew Acceptab6e Not Acceptable Needed Date inMietor: Neverro Mark G O O tiet297 VT Lead: Ryan. Thomes J VT Mgr: Schopfer Don K O O O 11/1297 1RC Chmn: Singh. Anand K D D 11/1497 O O O iiri7/97 m
SL Comments:
Prtnted 11/18,97 4 28.07 PM Pege 2 of 2
Northeast Utipties ICAVP DR NO. DR MP3 0470 Millstone Unit 3 Discrepancy Report Rev6ew Group: Operstone & Maintenerwe and Testang DR RESOLUTION ACCEPTED Rev6ew Element: Operehng Procedure , ,
1 Dieciphne: Opereuon.
D6*crepancy Type: 0 & M & T Procedure Om 4g
~
systenVProcess: $WP NRC sagrdacance level: 4 Date faxed to NU: I Date Published: 107197 Diecrepecy: Procedures not in place to ensure compliance with Plant Technical Specifications. j Deecription: The Technical Specifications require inspection of snubbers !
when their system has experienced unexpected, potentially )
damaging transients. The necessary procedures to ensure compliance with this Tech. Spec. were not in place at the time of this independent review. While this DR is written against SWP, j it does Apply to all systems.
Technical Specification 3/4.7.10d (page 3/4 7 23) states, 'An inspection shall be perf9rmed of all snubbers attached to !
sections of systems that have experienced unexpected, potentially damaging transients as determined from a review of I operational data and a visualinspection of the systems within 6 monthl,Iolbwing such an event. In addition to satisfying the i visualinspection acceptance criteria, freedom of motion of l mechanical snubbers shall be verified using at least one of tho ;
following: (1) manually induced snubber movement; or (2) '
i evaluation of in-place snubber piston setting; or (3) stroking the mechanical snubber through its full range of travel.'
S&L Request for It) formation, RFl Number MP3 219 was submitted on 7/18/97 requesting the Snubber Procedure used for Transient Event inspection to satisfy MP 3 Tech. Spec. 4.7.10d, page 3/4 7 23.
IRF Response ID : M3 lRF 00202 stated " Items 8 and 9 are not addressed by an existing procedure. AR 97019941 requires a new procedure or reviss 1 procedure will address these items at a later date".
Action Request, AR 97019941 was issued on 8/11/97 stating,
- Develop appropriate trigger to ensure compliance with Technical Specification 4.7.10d which requires snubbers to be examined following unexpected, potentially damaging system transients within 6 months of the event".
AR 97019941 was closed per Assignment completion Notes ,
'9/26/97: This assignment is closed to AR 97023549 and AR 97023551 which were generated for OPS Department to incorporate trigger into procedure AOP 3570 (Earthquake) and procedure OP 3263 (Unit 3 Duty Officer Actions Following Trip or ESF Activation) to prov'de Engineering notificat on to ev#uate this applicability of Tech. Spec. 4.7.10d following an event".
Procedures necessary to demonstrate how Millstone Unit 3 personnel satisfy Technical Specification 3/4 7.10d were not in nint n ne nf thm if' AVP pef ahlithAdfeulou dnf a nf LAnv 97 1997 Printed 11titL97 4 27.00 PM Pope 1 of 2
been UtWties ICAVP DR NO. DR MP3-0470 Millstone Unk 3 Ciscrepancy Ra,. ort Review vad invad we.d.d De inntator: Prver, W. O O O 551*S7 VT Leed: Beu, Ken O O O '*2*S7 VT Mgri Schopfer, Don K O O O io/15S7 IRC Ctann: Singn, Anand K Q Q Q 10/1&97 D*:
wvAuo:
DW: 11/13/97 .
McSOLUTION: NU has concluded that Discrepancy Report DR MP3 0470, has identified a condition previously discovered by NU which required correction. The attached memos from Nuclear Licensing and Tech Support maintain the position that NU had previously identified the discrepancy with satisfying Tech Spec surveillance requirement 4.7.10.d prior to the 5/27/97 date of ready for ICAVP.
Prev 6ously identified by NU7 #1 vee O No Non Descrepent CondM60n?O Yee io) No Resolut6on Pend 6ng?O vee *) No Resolut6onUnresolved?O vos *> No Review Acceptable Not Acceptable Needed Date ggg P W.
VT t.ead: Base, Ken b b VT Mgr: Schopfer, Don K IRC Chenn: sngh, Anend K D*: 11/13/97 SL comments: The NU memo NL 97 049 dated 2/26/97, does identify the Technical Specification 4.7.10d as requiring review prior to the 5/27/97 ICAVP established review date. If this information had been sent with the original Request for Information (RFl MP3-219), this Discrepancy Report would not have been generateo.
Eyed 11/1&S7 4 27.17 PM Page 2 of 2
No'theast Utilities ICAVP DR No. DR MP3 0180 Milletone Unit 3 Discrepancy Report Rev6ew Oroup: Programmetc DR RESOLUTION REJECTED Review Element: Change Process g g,,
p Diecipline: Otho' D6ecrepency Type: Uceneng Document Ow
- 4) No
~
systemtrocess: N/A NNc Signiacance levot: 4 Date faxed to NU:
Date Published: &W97 06ecrepency: TechnO. Justification for Changes to FSAR Descr6puon: The Final Safety Analysis Report Change Request (FSAECR) packages discussed below did not contain sufficient documented technical justification for the changes to FSAR.
Discussion:
The following FSARCR packages have been reviewed and it is considered that documentation contained in the packages is not sufficient to determine if the changes are technically acceptable, in all cases, the change package contained no 50.59 safety evaiustion. The absence of a safety evaluation in does not mean that the change was not technically acceptable, howevur lack of such documentation and the associated evidence that various technical and regulatory issues were considered in making the change, renders the technical acceptability of the changes as indeterminate within the scope of review of the IC/NP. A discussion of each of the four (4) FSARCRs follows.
- 1. FSARCR 95-MP316 An independent Safety Engineering Group (ISEG) report E9?-
010 identified in part, certain discrepancies in the UFSAR when compared to station procedures and practices. To resolve these discrepancies, the FSARCR changed the FSAR to remove a monthly check of the
- chilled water control valves for movement and leaks .' The FSARCR also changed a monthly testing frequency to quarterly for the Conti01 Building air conditioning and ventilation systems not normally in use (standby) to be consistent with station practices. The FSARCR presents a brief justification for the deletions which states how valve mtvements and leaks are checked. However, the decision was apparently made not to descnbe such checks in the FSAR. The justification presented for changing the testing frequency from monthly to quarterly is to reflect the current station practice. There is also a statement 'No safety evaluation required' without supponing explanation. The FSARCR implements changes to procedures described in the SAR and therefore should have had a safety evaluation or refered to one performed for a corresponding proceduro change (n), as arolicable. The package does not document what the requirements are for testing from a design basis and/or regulatory compliance perspective. There is no evidence that other controlling documents such as SEris or licensing comittmente were reviewed nor other documentation to definitively conclude (as stated on the form), that 'certain information in the FSAR....has been rendered !ncorrect...'
Printed 11/1M;7 4 M.33 PM FSARGR ^" """" "' '
Page 1 of 4
0
+
DR No, DR MP3 0140 Northeast Utilities ICAVP Milletone Unit 3 Discrepancy Report A safety evaluation is not included in the package. Technical ,
acceptability must consider the design and licensing basis requirements (not stated in the package) compared to the FSAR ,
changes being imple.nented, it is not sufficient to state Viat the changes are justified by the actual as-built condition and/or cu Tent surveillance practices of the plant without stating or providing other eviderice that the as-built condition / surveillance practices are in accordance with design and licensing basis requirements.
- 3. FSARCR 89-MP3-9 Same explanation as item No. 2 above.
- 4. FSARCR 87 MP310 Same explanation as lem No. 2 above.
Review Vehd inval6d Needed Date S'15S7 initletort Newtro. Mark O O O O 8'55S7 VT Lead: Ryan. Thomas J O O VT Mgr: Schopfer. Don K O Q Q &%97 INc Chmn: Singh, Anand K Q Q Q S'2s/37 Dese:
INVAUD:
Dele: 11/12/97 RESOLUTION: Disposition NU has concluded that Discrepancy Report, DR MP3-0180, has identified a condition previously discovered by NU which requires correction. Referto the attached memo, NL-96-263 dated 12/11/96, for an explanation of the review of historical FSARCRs for Safety Evaluations. In short,525 historical FSARCRs from 1986 to 1996 were reviewed to determine if any were processed without a Safety Evaluation when one was needed, This review identified 80 FSARCRs that rec'ti-1 formal screening in accordance with NGP 4.03 Fin. 2, Stand Alone FSARCR 10CFR50.59 Safety Evaluation r .11ng Form. Three of the four FSARCRs identified in the OR were among the 80 selected for screening (see attached copies of NGP 4.03 Figure 7.6.) FSARCR 95-MP3-16 was determined to require a Safety Evaluation. Currently, the Safety Evaluation I
has not been completed. FSARCR 92-MP3 35 was determined not to require a Safety Evaluation based on the editorial nature of the changM. FSARCR 89 MP3 9 was determined to require a l Safety Evaluation. The Safety Evaluation (attached) has been l complete <1 and signed off.FSARCR 87-MP310 was not among i the 80 changes selected for screening. This is because the l changes made were considered clarifications and did not meet I the enteria established at that time for FSARCRs requiring l screening (refer to attached memo NL 96-263 dated 12/11/96, for selection criteria). Based on the program NU had established l Printed 1in&W 4 29 do PM Page 2 of 4 l
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Northeast UtiHtles ICAVP DR No. DR MP3 0180 )
Millstone Unk 3 Discrepancy Report l st that time, the FSARCR did not require a screening for a Safet> Evaluation. No further screening will be performed on the FSARCR since it met the original screening criteria.The technical review is documented in Section 2 of each FSAR CR.
Conclusion NU has concluded that Discrepancy Report. DR MP3-0180, has identified a condition previously disociered by NU which i requires correction. During the 50.54(f) effort 80 historical i FSARCRs were chosen based on specified criteria for Safety l Evaluation screening. FSARCRs 95 MP316,92 MP3 35 and 89 MP3 9 were among the 80 identified for screening. The associated screening forms and Safety Evaluation are attarted. )
87 MP310 was not among the 80 chosen for screening because it did not meet the established selection criteria. Based on the program NU had established at that time, the FSARCR did not require a scraening for a Safety Evaluation. No further screening will be performed on the FSARCR since it met the original screening criteria.
Prev 6ously idenuned by Nu? O Yes #1 No Non D6ecrepent condition?Q Yes @ No ResolutionPending?O vos
- No ResolutionUnresolved?O vos
- No Review
- g. AccepL$le Not Acceptable Needed Date
,,,g VT Lead: Ryan, Thomme J VT Mgr: schopfer Don K IRC Chmn: singh, Anand K oste: 11/12/97 st connente: Regarding FSARCR 87 MP310:
While most of the changes in FSARCR 87 MP310 are clarifications, there is a change in FSARCR 87 MP3-10 which should not be character 1 zed as a " clarification" since the change affects the frequency of testing for the MSIVs from "at least montbly" to ' quarte'1y* .
l This particular change constlutes changes to procedures desenbed in the FSAR and t.hould have a safety evaluation in accordance with NU's screening criteria. This is consistent with the rationale and conclusion that a safety evalution was required l In stand alone FSATICR 10CFR50.59 safety evaluation SCREENING form for FSARCR V95 MP316 dated 2/28/97 for which a copy was provioed in response to this DR. Resolution of this issue should also address whether NU's criteria for determining that a safety evaluation was required, was consistently applied in the review process which looked at 525 historical FSARCRs.
Regarding FS ARCR MP3-95-MP3-16:
l As noted in the Licensee's response to this DR, the stand alone I screening dated 2/28/97 which was performed for FSARCR 95-MP3-16 determined that a 50.59 evaluation was required. Eight (8) months has elapsed since this determination and the Licensee Pnnted 11n&97 4 2945 PM Page 3 of 4 i
e ICAVP DR No. DR MP34180 Northeast Utilities Millstone Unit 3 Discrepancy Report reports in response to this DR (without explsnation) that the 50.59 l has yet to be completed. in the absence of further information, it l I
is considered the timeliness in completing such 50.59 evaluation i
is not acceptable. The following questions should be addressed' Was this an isolated case of an oversight, or is there a concem !
regarding the completion of the 50.59 screenings for the other FSARCRs which also resulted in a determination that a 50.59 was 4 required? For those screenings determined to require a safety j evaluation, how is the completion of the safety evaluatlor; being j tracked and is completion required pior to startup?
PMed 11MlVS7 4 29 47 PM Page 4 or 4
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Northeast Utilities ICAVP DR No. DR MP3-0204 wilstone Unit 3 Discrepancy Report Rev6ew Grt up: Programmetc DR RESOLUTION REJECTED Reh EW: Change Process Peterenal Operatety 14eue D6ecip46ne: Piping Desig' Discrepency Type: Corrective Action Om e) No SystemProcess: N/A NRC Sign 6Aconce level: 4 Date faked to NU:
Date Published: 9/29/97 bacropency: Incomplete Nonconformance Report Package Descispt6on: 1. NCR # 3-910018 deals with weld inoications on the travelling screen wash piping welds. The pipng had pin hole leaks. In an attempt to remove the defects, a cavity of approximately 6 inches by 3 inches was opened in the piping. The disposition was to weld repair the cavity without deterrnining the extent of the remaining 6inear indications and porosity. The NCR states:
'The defects probably exist throughout the circumference of the welds."
Quality Control was to note any remaining indicatinns after weld repair for information only. The disposition states that the weld defects were caused by the manufacturer and that the piping had been in service for approximately three years. No further justification for the dispositic.7 was provided.
- 2. No safety evaluation, environmental evaluation, safety evaluation screening or environmental screening was found in the package.
- 3. Parts of the NCR package are difficult to read. This does not meet the requirements of NU's commitment to NRC Regulatory Guide 1.88.
Review Valid invaled Needed Date inn 6etor: shoppard, R. P. O O O S' 5S7 VT Lead: Ryon. Thomas J Q Q Q 9/1597 VT Mgn schopfer, Don K O O O S'nS7 IRC Chmn: Singh. Anand K O O O W2SS7 o ee:
INVAllD:
Date: 11/7/g7 RESOLUTION: Disposition:
NU has concluded that DR MP3-0204 does not represent a discrepant condition.
- 1) The dL taosition of the NCR states that " leaving the remaining linear indications will not detrimentally affect the structural integrity of the welded joint". No fudher technicaljustification is required in this case.
- 2) The nonconformance procedure at the time Of the NCR did not require an environmental or safety screening,but required a safety evaluation if necessary. A safety evaluation was not Printed 11/1&97 4 3o 21 PM Page 1 of 3
o e'
DR No, DR MP3 02C4 Northeast Utilities ICAVP Millstone UnN 3 Discrepancy Report required as Ind ated on NCR form NEO 3.05, Rev 1. The proper evaluation was performed in accordance with applicable procedures.
- 3) Reg. Guide 1.88 concems storage and maintenance of records. The copy of the NCR reviewed is stored and maintained in NDS records on microfilm. Any reduction in legibility of the hard copy provided is due to the reproduction process. The NCR on microfilm is both legible and maintained in accordance with the regelrements of Reg Guide 1.88. A legib;e copy of NCR 391018 is being transmitted with this IRF, Significance level criteria du not apply as this is not a discrepant condition.
ConcluCon.
NU has concluded that Discrepancy Report DR MP3-0204 does not represent a dissepant condition. As detailed in the disposition, adequate technical justification was provided in this NCR, a proper evaluation of this NCR was performed in accordance w'th applicable procedures, and NU's records comply with Reg Guide 1.80. Significance level criteria do not apply as this is not a discrepant condition.
~
Prev 6ously klontened by NuIO Yes (#1 No Non Descrepent Corulet60n?Q Yes @ No Rooolut6on Pending?O Yes 5) No Resolut6on Unresolved?O Yo. 4) No Review inilleton sheppard, R. P.
VT Lead: Ryan. Thomeo J VT Mgr; schcpfer, Don K 1RC Chrnn: sirgh. Anand K Dato 11/7/97 sL conwnents: 1. The resolution of item 3 of this discrepancy is acceptable.
- 2. The visual weld inspection plan lists ANSI B31.1 as the design code for this piping. Paragraph 127.4.11 of ANSI B31.11973 requires that the type of examination and the limits of imperfections for repair welds shall be the same as for the original weld. Paragraph 136.4.2 of B31.1 requires that visual examination consist of observation of whatever portions of a component or weld are exposed to such observation, either before, during or after fabrication, it further regelres welds and piping have no extomal cracks or meet the limitations of the piping material specification. A 312 72a, section 12, requires that pipes be reasonably free from injurious defects.
Leaving the linear indications in the pipe without repair appears to be a design change. Section 6.2.2 of NEO 3.05, Revision 1, states that a disposition to correct an inadequate design should include a requireroent to change the design or the NCR should reference a design change document.
, 'o s*
DR No, DR44P3 020#
Northeast Utilities ICAVP Millstone UnM 3 Discrepancy Report evaluation was not required. Possibly the reference is to a *SSH Evaluation" not being required. Substantial Safety Hazards (SSH) are defined in 10 CFR 21.3 and the evaluation to determins whether a deviation or failure to comply could cause or be assodated with a SSH is not riscessarily related to safety evaluations. If in fact the design was changed by this NCR, a design change should have been processed which would have evaluated the safety and environmentalimplications of the change.
NU is requested to provide further information justifying the adequacy of the pipe and its compliance with the licensing bases.
PrWed 11/1&97 4 30 31 PM Page 3 of 3
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