ML20216B701
| ML20216B701 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 03/06/1998 |
| From: | Grobe J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Patulski S WISCONSIN ELECTRIC POWER CO. |
| References | |
| 50-266-97-10, 50-301-97-10, NUDOCS 9803130158 | |
| Download: ML20216B701 (2) | |
See also: IR 05000266/1997010
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- March 6, 1998 m\/LOi\QCD
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Mr. S. A. Patuiski
Site Vice President _
Point Beach Nuclear Plant
6610 Nuclear Road
Two Rivers, WI 54241
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SUBJECT: NOTICE OF VIOLATION (NRC INSPECTION REPORT NO. 50-
266/97010(DRS); 50-301/97010(DRS)
Dear Mr. Patulski:
This will acknowledge receipt of your letters dated October 15,1997 and February 11,
1998, in response to our letter dated September 15,199 , transmitting a Notice of Violation
associated with the above mentioned inspection report at the Point Beach Nuclear Plant. We
have reviewed your corrective actions and have no further questions at this time. These
corrective actions will be examined during future inspections.
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, Sincerely,
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/s/ Harold O. Christensen
l John A. Grobe, Director
Division of Reactor Safety
Docket No. 50-266
Docket No. 50-301
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Enclosures: 1. Ltr 10/15/98, S. Patutski, WEPC, to
US NRC
! 2. Ltr 2/11/98, S. Patulski, WEPC to
US NRC f
Sge Attached Distribution /
DOCUMENT NAME: G:DRS\ pol 030_8.DRS
To ,eceive a copy of this documefA, Indicate in the bom! "C" e Copy without attachment / enclosure "E" = Copy with attachment /encionure "N* a No copy
l OFFICE Rlli lV Rlilj / Rill l 7) Rlli lQ
NAME Butler:sd j)/13 GliRula /fffF Gardner PJV Grobe @'
DATE 03/6 /98 03///98 03/t, /98 03/6/98 '
OFFICIAL RECORD COPY
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9803130158 980306 I
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S. Patuiski! 2 March 6, 1998
cc w/encis: R. R. Grigg, President and Chief
Operating Officer, WEPCO
A. J. Cayla, Plant Manager
B. D. Burks, P.E., Director -
Bureau of Field Operations
Cheryl L. Parrino, Chairman
Wisconsin Public Service
Commission
State Liaison Officer
Distribution:
Docket File w/encls SRI Point Beach w/encls ~
PUBLIC IE-01 w/encls Rill Enf. Coordinator w/encls
A. B. Beach w/encls IEO w/encls (E-mail)
Project Manager, NRR, w/encls DRS (2) w/encls
Rill PRR w/encls DOCDESK (E-mail)
.TSS w/encls GREENS
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! Electnc
POWER COMPANY
Point Beoct Nuclear Plant (414) 755-2321
6610 Nuclear Rd.. Two Rivers. WI 54241
NPL 97-0648
October 15,1997
Document Control Desk
U. S. NUCLEAR REGULATORY COMMISSION
Mail Station PI-137
Washington, DC 20555
Ladies / Gentlemen:
DOCKETS 50-266 AND 50-301
RESPONSE TO A NOTICE OF VIOLATION AND APPARENT VIOLATIONS
IN NRC INSPECTION REPORTS 50-266/97010 (DRS) AND 50-301/97010 (DRS)
POINT BEACil NUCLEAR PLANT, UNITS 1 AND 2
In a ietter from Mr. John A. Grobe dated September 15,1997, the Nuclear Regulatory
Commission forwarded the results of an inspection conducted by your stafTat our Point Beach
Nuclear Plant from May 19,1997, through June 13,1997. This inspection report included a
Notice of Violation that identified two violations ofNRC requirements This inspection report
also included a description of four apparent violations that are being considered for escalated
enforcement action.
As noted in the NRC's September 15,1997, letter, the four apparent violations were identified by
Wisconsin Electric and were appropriately reported to the NRC. These issues have been
previously presented to the NRC during senior management meetings and discussed during the
June 13,1997, inspection exit meeting. We endorsed the NRC's belief that a pre-decisional
enforcement conference was not necessary for these four apparent violations. As such, we
notified Mr. James W. McCormick-Barger of NRC Region III on September 23,1997, of our
intention to respond to the four apparent violations within 30 days ofyour September 15,1997,
letter.
We have reviewed the Notice of Violation and, pursuant to the provisions of 10 CFR 2.201, have
prepared a written response to the two cited violations that is included as Attachment A.
Attachment B to this letter provides our response to the four apparent violations.
We believe the attached reply is responsive to the Notice of Violation and fulfills the requirements
identified in your September 15,1997, letter. New, previously undocketed commitments in this
response are identified by italics.
OCT !. O M
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Ifyou have any questions or require additional information regarding this response, please contact
me.
Sincerely,
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ott A. Patulski
Site Vice President
Subscribed and sworn before me on
this _fd_ d =~ ,f W .~ 1997.
&0 .Wb
Notary Pdblic, State of Wisconsin
My commission expires h A.soe / .
Attachment
FAF/
cc: NRC Regional Administrator l
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NRC Resident Inspector
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ATTACHMENT A
DOCKETS 50-266 AND 50-301
RESPONSE TO NOTICE OF VIOLATION l
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IN INSPECTION REPORTS 50-266/97010 (DRS): 50-301/97010 (DRS)
POINT BE.ACH NUCLEAR PLANT, UNITS 1 AND 2
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During an NRC inspection conducted from May 19,1997, through June 13,1997, two violations !
of NRC requirements were identified. Inspection Report 50-266/97010 and 50-301/97010 and
the Notice of Violation (Notice) transmitted to Wisconsin Electric on September 15,1997,
provide details regarding the violations.
In accordance with the instructions provided in the Notice, our reply to the violations includes:
(1) the reason for the violat an, or if contested, the basis for disputing the violation; (2) the
corrective action taken and thm results achieved;(3) corrective action to be taken to avoid further
violations; and (4) the date wn. i full compliance will be achieved.
Violation 1:
"10 CFR 50, Appendix B, Criterion V, " Instructions, Procedures, and Drawings," requires, in
part, that activities affecting quality be prescribed by documented procedures of a type l
appropriate to the circumstances and shall be accomplished in accordance with these procedures.
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Nuclear Procedure (NP) 5.3.1, " Condition Reporting System," states, in part, that the
condition report (CR) is a method to identify, evaluate if necessary, and correct an event i
or condition which has the potential to adversely affect operability of safety-related
systems or equipment including discrepancies associated with calibrations.
Contrary to the above, the inspectors identified on June 5,1997, that the licensee failed to
follow NP 5.3.1. A condition report was not initiated for the G03 emergency diesel
generator (EDG) voltage monitoring relay out-of-calibration condition. The relay was
! successfully re-calibrated on May 22,1997; however, the out-of-tolerance condition was
not evaluated as to the relay's potential impact on past EDG operability.
This is a Severity Level IV violation (Supplement I)."
Reason for Violation 1:
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We concur with Violation I as characterized in the inspection report. The reason the violation
occurred was because the electrical maintenance supervisor in charge of the work misinterpreted
the condition reporting procedure, NP 5.3.1, to mean that if an "as-left" condition was found to
be out of tolerance, then a Condition Report should be written. When your inspector brought the
item to his attention, he realized that he was expected to document the "as-found" out of
calibration tolerance via a Condition Report. Initiation of the Condition Report would have
driven the documented review of the "as-found" condition for operability
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Corrective Action Taken:
1. Condition Report CR 97-1779 was initiated on June 5,1997,
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2. A discussion of this event with electrical and I&C maintenance supervisors determined l
that this was an isolated case. Other supervisors were aware of the expectation to initiate
a condition report when an "as found" out of tolerance condition is identified that has the
potential to affect safety-related equipment.
3. A memo was issued to all electrical supervisors on June 20,1997, reiterating ,
management's policy and expectations regarding the actions to be taken when an out of 1
tolerance condition is identified. On June 21,1997, electricians were briefed on this
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4. A review of the out of tolerance relay condition concluded that a past operability
determination was not required.
Corrective Actions Taken to Prevent Recurrence:
Maintenance Administration Manual instruction MTAM 3.8, " Submittal of Condition Reports for
Out of Tolerance Readings," was issued on October 13,1997. This document provides formal
guidance to ensure that appropriate actions are initiated and reviews are conducted to assure
eugipment operability and reliability.
Date Full Compliance Will he Achieved:
Full compliance with NRC requirements was achieved on October 13,1997, upon issuance of
formal guidance for initiation of Condition Reports for out of tolerance readings.
Violation 2:
"10 CFR 50, Appendix B, Criterion IX, " Control of Special Processes," requires, in part, that
measures shall be established to assure that special processes, including welding, are controlled
and accomplished by qualified personnel using qualified proc'dures in accordance with applicable
codes, and standards.
ASME Section IX,1995 Edition, requirement QW-409.1, as implemented by requirement
QW-256, requires that "A change in the type of current or polarity, an increase in heat input, or
an increase in volume of weld mett.1 deposited per unit length of weld, [is not allowed] over that
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qualified.
Contrary to the above, the inspectors identified on June 13,1997, that weld procedure WPS
GT-SM/3.3-2 PB, Revision 1, did not meet the requirements of QW-409.1. The weld procedure
allowed welding to be performed with an increase in heat input over that which had been
r demonstrated through Charpy impact testing of the qualification weld.
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allowed welding to be performed with an increase in heat input over that which had been
demonstrated through Charpy impact testing of the qualification weld.
This is a Severity Level IV violation (Supplement 1)."
Response for Violation 2:
Wisconsin Electric has reviewed welding procedure specification (WPS) GT-SM/3.3-2 PB and its
supporting procedure qualification record (PQR) GT-SM/3.3-Q2 for conformance to the
requirements of the 1995 edition of ASME Section IX and the requirements of the 1986 edition
of ASME Section III, Subsection NB, regarding establishing the heat input as required by
QW-409.1. It is Wisconsin Electric's opinion that the heat input dats recorded on PQR
GT-SM/3.3 Q2 is consistent with the philosophy of and meets the requirements of the previously
stated editions of ASME Section IX and Section III, Subsection NB. Therefore, we do not
believe that a violation of 10 CFR 50, Appendix B, Criterion IX occurred. Our basis for disputing
the violation follows.
The welding procedure requirements for the girth weld are specified in ASME Section III,
1986 Edition, Article NB-4330. The area ofinterest relating to impact testing requirements is
identified in Article NB-4335, which states, "The weld procedure qualification impact test
specimens shall be prepared and tested in accordance with the applicable requirements of
NB-2330 and NB-4334." I
These two articles dictate the requirements for the preparation of the weld specimen. All ,
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requirements identified in these sections for the preparation of the weld specimen were met in
their entirety in PQR GT-SM/3.3-Q2. Please note that these articles do not address any variation
in heat input among multiple passes. Furthermore, these sections also specify the location of the
i impact test coupon and reference ASME Section II SFA-5.1 for details on coupon location. The
location of the impact test coupon is determinea by test specimen geometry. No reference is
made in any of the above referenced articles to locate the test coupon in relation to heat input of
l individual passes. The test coupon removed to support PQR GT-SM/3.3-Q2 was located in
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compliance with the referenced articles.
Therefore, the preparation of the test specimen and the location of the impact test coupon in
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support of PQR GT-SM/3.3-Q2 was in compliance with the ASME Boiler and Pressure Vessel
Code.
The Notice of Violation referencesSection IX QW-409.1 which states that, "An increase in heat
input, or an increase in volume of weld metal deposited per length of weld, [is not allowed] over
that qualified." Please note that QW-409.1 does not specify whether the area in question is the
weld specimen or impact test coupon. The requirements of QW-409.1 were met in the
preparation of the weld specimen for PQR GT-SM/3.3-Q2 in that the maximum heat input used
on an individual pass on the test weld specimen was appropriately identified on the PQR.
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The Notice does not specifically reference Code Interpretation IX-92-69, however, the verbiage
used in the Notice and discussions with NRC inspectors on this issue are consistent with
IX-92-69. This interpretation attempted to address the issue questioned by the NRC, i.e., the i
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location of the impact test coupon in relation to the maximum heat input used in the test
specimen.
IX-92-69 was initially received by ASME in January,1992, and various parts were answered over
the next several Code meetings. The main issue of controversy that remained unresolved was
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regarding location ofimpact test specimens relative to the various heat inputs that might occur at
various locations in a test coupon joint.
In the May,1993, meeting, the inquiry was approved by the Section IX Subcommittee, subject to
the approval by the Main Committee. At the same meeting, proposed changes to ASME
Section IX, QW-409.1 were also approved by the Section IX subcommittee. These changes 4
would have changed the article to read as follows:
"A change in the type of current or polarity, an increase in heat input, or an increase in
volume of weld metal deposited per unit length of weld, over that qualified. The
maximu n heat input or volume of weld metal per unit length of weld quaLijed shall be
determined from the weld passes with the nominal heat input or volume of weld metal per
unit length or weld sampled by the impact test specimens."
Note that this wording is consistent with the alleged violation. However, this interpretation
generated significant controversy within both the ASME Section IX Subcommittee and the Main
Committee. Although the interpretation and the proposed change were passed by the Section IX
Subcommittee, three negative votes were noted, based primarily on the practical implementation
of the proposed changes. Furthermore, the Main Committee registered fifteen negative votes, and
both the interpretation and the proposed changes to QW-409.1 were voted down at the Main
Committee level.
Because the interpretation was voted down at the Main Committee level, the interpretation should
not have been published. Many older interpretations were not well received by the Main
Committee and should not be accepted carte blanche, thus supporting what we have understood
to be the NRC's position of not endorsing Code interpretations. Shortly after this interpretation
was published, ASME changed their rules of operation such that interpretations are not published
until the Main Committee has approved both the interpretation and any corresponding changes to
the Code. Efrorts are currently underway to review outstanding interpretations, and either
incorporate or withdraw the interpretations. In fact, Interpretation IX-92-69 was recently
reviewed during a September 16,1997, meeting of the ASME Section IX Subcommittee. It is
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our understanding that the Section IX Subcommittee voted unanimously to withdraw the inquiry I
and the December Addenda should reflect the oflicial withdrawal ofInterpretation IX 92-69.
Wisconsin Electric agrees that, if Code Interpretation IX-92-69 had been incorporated into
ASME Section IX QW-409.1, the violation as stated may be valid. However, the interpretation
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was not incorporated, and in fact, has been withdrawn altogether. Therefore, no violation of
ASME Code requirements occurred.
Wisconsin Electric acknowledged the controversy associated with this issue, and in addition to
several open discussions with NRC inspectors, we took the following additional conservative
actions to facilitate closure of this issue. Both Wisconsin Electric and our contractor, SGT Ltd.
fully believe that the girth weld, as originally qualified, was fully Code compliant. At our request,
SGT Ltd. undertook to repeat the PQR, assuring that the heat input noted in the original PQR
was boundea within the impact test coupon of the second PQR. The ,epeat of this PQR and the
associated testing verified that WPS GT-SM/3.3/3-2 PB is qualified. Therefore, there is no
question regarding the Code compliance of the replacement steam generators as installed.
Furthermore, two acknowledged ASME Code experts were consulted on this issue. Both
consultants arrived at the same conclusion; namely, that the PQR as originally performed, was in
full compliance with the ASME Boiler and Pressure Vessel Code Funher, as a result of this
review on this issue and with the intent to bring closure to this issue, one of the consultants
generated the request for additional review and subsequent recent withdrawal of Code
Interpretation IX-E2-69 by the ASME Section IX Subcommittee.
Wisconsin Electric believes that the original PQR GT-SM/3.3-Q2 was in full compliance with the
ASME Boiler and Press 9re Vessel Code. No additional actions were or are required to achieve
compliance.
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ATTACHMENT B
DOCKETS 50-266 AND 50-301
RESPONSE TO APPARENT VIOLATIONS
IN INSPECTION REPORTS 50-266/97010 (DRS): 50-301/97010 (DRS)
POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2
During an inspection conducted between April 29,1997, and June 12,1997, four apparent
violations related to 10 CFR 50 Appendix R requirements vmre identified. The apparent
violations involve two Licensee Event Reports and associated condition reports that detail the
Appendix R requirements that have not been met. Inspection Report 50-266(301)/97010
provides details regarding the violations.
We have accepted the opportunity to respond in writing to the apparent violations, as opposed to
requesting a predecisional enforcement conference. Therefore, our written response was prepared
in accordance with the instructions provided in the Inspection Report, and includes: (1) the
reason for the apparent violation, or if contested, the basis for disputing the apparent violation; (2)
the corrective steps that have been taken and the results achieved;(3) the corrective steps that will
be taken to avoid further violations; and (4) the date when full compliance will be achieved.
Almarent Violation:
1. " Appendix R, Section llI.G.2 requires, in part, where cables that could prevent operation or
cause maloperation due to hot shorts, or shorts to ground, of redundant trains of systems
necessary to achieve and maintain hot shutdown conditions are loca'r i within the same fire
area, a means to ensure one of the redundant trains is free of fire dav 4e shall be provided.
The following are examples of an apparent violation of 10 CFR 50 Appendix R
Section Ill.G.2 requirements:"
a. "The licensee failed to identify during the original Appendix R reviews the lack of
125Vdc circuit breaker electrical coordination as described in CR 96-136. The cables
routed through conduit D04-7 were susceptible to fire induced shorts for a fire in the
North auxiliary feedwater area. This had the potential to cause the " yellow" instrument
channels to be loss (sic). The " yellow" instrumer.:s were the redundant train of safe
shutdown components used in this fire area."
b. "The licensee failed to identify during the original Appendix R reviews the lack of
125 Vdc circuit breaker electrical coordination as described in CR 96-889. The cables
routed through conduit D04-7 were susceptible to fire induced shorts for a fire in the
' North auxiliary feedwater area. This had the potential to cause the loss of all EDGs.
The G02 EDG was the analyzed redundant train of safe shutdown equipment used in
this fire area."
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Wisconsin Electric Response to Apparent Violation
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We agree that the violations occurred as described. These violations were identified by our staff
in Condition Reports CR 96-136 and CR 96-889 and reported to the NRC in Licensee Event
Report (LER) 266/97-020-00. This LER described the reason for the violations and the related
corrective actions. Those reasons and corrective actions from the LER are summarized below i
and supplemented with new information as necessary. Please note that a supplement to that LER
was issued on October 14,1997.
Apparent Violation 1.a
Reason for Apparent Violation:
As described in the LER, this condition was caused by inadequate design review of the EDG
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Addition Project (Modification Request MR 91-116). The modification routed a new " associated
circuit" through the AFW Pump Room without verifying appropriate bre& coordination.
Therefore, this modification introduced a new failure mechanism for the potential loss of the
yellow instrument channel for certain fires.
Corrective Actions Taken:
As described in the LER, a shon-term compensatory measure was established to reduce the l
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likelihood and severity of a postulated fire in the North Zone of the AFW Pump Room until
permanent modifications could be installed. During the summer of 1997, modifications and
procedure changes were completed to remedy the condition. Modification Request MR 97-058
was completed to install a 1-hour rated fire wrap on conduit D04-7 and junction box JD4176 in
the AFW Pump Room. Installation of the fire wrap provided separation in accordance with
Appendix R III.G.2.c for a fire in the AFW Pump Room and thereby, provided protection from a
potential fault on cable ZBD0408 A and the consequential loss of the " yellow" instrument channel.
To address the inadequate design review which contributed to the mis-routed cable, we have
directed the Fire Protection Engineer to review modificationsforfire protection and Appendix R
implicationsprior to installation. This is an interim measure until the design controlprocess is
reviewedas described below.
Corrective Actions to be Taken: ,
The Appendix R Rebaselining Project will address the programmatic failures which caused the
apparent violations identified in the Inspection Report, specifically including a review of the
design control process with respect to fire protection and Appendix R. We expect that the
revised design control process will make available more of the resources and experience that will
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ensure the plant design continues to comply with fire protection and Appendix R requirements.
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Annarent Violation 1.b
Reason for Apparent Violation:
This condition was caused by inadequate design review of the EDG Addition Project
(Modification Request MR 91-116). The design should have ensured that the DC control power
for at least one emergency diesel generator was protected from a fire in the North Zone of the
AFW Pump Room.
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Corrective Actions Taken:
As described in the LER, a short-term compensatory measure was established to reduce the
likelihood and severity of a postulated fire in the North Zone of the AFW Pump Room until
permanent modifications or procedure changes could be installed to remedy the condition.
During the summer of 1997, circuit analyses conclu,ded that the gas turbine generator (G-05)
would be the appropriate power source for safe shutdown in the event of a fire in the north halfof
the AFW Pump Room. On July 1,1997, Fire Emergency Plan (FEP) 4.12, " Auxiliary Feedwater
Pump and Vital Switchgear Area" was revised to direct operators to align G-05 for a fire in the
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north area of the AFW Pump Room. The selection of this power source obviates the use of
emergency diesel .senerators for this particular fire scenario.
As discussed above for Apparent Violation 1.a. an interim measure has been taken to remedy the
design control error which contributed to this apparent violation. l
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Corrective Actions to be Taken:
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As discussed above for Apparent Violation 1.a, the Appendix R Rebaselining Prcject will address
the programmatic failures which caused the apparent violations identified in the Inspection Report
and will 1.pecifically address the design control process with respect to fire protection and ;
Appendix R. i
Date When Full Compliance Will be Achieved:
PBNP was in compliance for these apparent violations when the aforementioned modifications
and procedure changes were completed in July,1997.
Apparent Violation 2:
" Appendix R,Section III.L.1 requires, in part, that alternative or dedicated shutdown capability
provided for a specific fire area be able to: (a) achieve and maintain subcritical reactivity
j . conditions in the reactor; (b) maintain reactor inventory; (c) achieve and maintain hot standby
' conditions; (d) achieve cold shutdown conditions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; and (e) maintain cold shutdown
conditions thereafler. The following are examples of an apparent violation of 10 CFR 50
Appendix R Section III.L.1 requirements:"
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l a. "The licensee failed to identify during the original Appendix R reviews that the original
safe shutdown analysis did not adequately document the capability of the G01/G02 EDGs
to start and flash their fields with their output breaker closed and 480 volt loads connected
to the EDG bus as described in CR 94-328. The potential existed for the alternative
shutdown path, use of the G01 or G02 EDG, to be loss (sic) leaving no success path for
b. "The licensee failed to identify during the origirial Appendix R reviews that the original
safe shutdown analysis did not adequately document the ventilation requirements for safe
shutdown equipment as described in CR 96-372. The potential existed for alternative safe
shutdown equipment to be loss (sic) due to the lack of room cooling."
Wisconsin Electric Response to Annarent Violations
We agree that the violations occurred as described. These violations were identified by our staff
in Condition Reports CR 94-328 and CR 92-372 and reported to the NRC in Licensee Event
Report (LER) 266/97-020-00. This LER described the reason for the violations and the related
corrective actions. Those reasons and corrective actions from the LER are summarized below
and supplemented with new information as necessary. Please note that a supplement to that LER
was issued on October 14,1997.
Apparent Violation 2.a
Reason for Annarent Violation:
This condition was caused by inadequate analysis of essential equipment operability requirements
during the original Appendix R Safe Shutdown Analysis. The original analysis did not adequately
document the capability of the EDG to operate under these conditions.
Corrective Actions Taken:
During the summer of 1997, plant modifications and procedure changes were completed to
remedy the condition. Modification Request MR 97-051 was installed to allow for circuit
isolation of gas turbine generator G-05 control circuits that traverse the Cable Spreading Room.
In addition, the abnormal operating procedure for postulated fires in the Cable Spreading Room
and Control Room (AOP-10A) was revised to align G-05 as the shutdown power source. This
procedure change obviated the use of emergency diesel generators G-01 and G-02 for these fire
scenarios. Therefore, the unanalyzed starting sequence described in the violation is no longer
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Corrective Actions to be Taken:
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The Appendix R Rebaselining Project will address the programmatic failures which caused the
apparent violations identified in the Inspection Report.
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Annarent Violation 2.b
l Reason for Annarent Violation:
This condition was caused by inadequate analysis of essential equipment operability requirements
during the original Appendix R Safe Shutdown Analysis. Ventilation equipment essential for safe
sbdown should have been protected from fire damage and its operability ensured or alternatives
provided for any consequential loss of ventilation, as required by Section III G of 10 CFR 50,
Appennix R.
Corret tive Actions Taken:
A calculation was prepared to determine room temperatures for areas containing safe shutdown
equipment. Fire areas where loss of ventilation could disable safe shutdown equipment were
identified, and the use ofportable fans after one hour was recommended for cenain fire areas.
Additional portable fans were purchased and are currently stered in appropriate plant locations.
In addition, plant procedures have been enhanced to provide guidance for supplemental cooling
for specific rooms.
Corrective Actions to be Taken:
As described in our letter dated August 27,1997, we plan to complete performance testing of
representative Appendix R ventilation equipment and develop qualitative criteria to verify the j
capability of all remaining configurations by October 31,1997. '
The Appendix R Rebaselining Project will address the programmatic failures which caused the
apparent violations identified in the Inspection Report. In addition, the Appendix R Rebaselining i
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Project will specifically evaluate the support requirements for safe shutdown equipment including
the environmental requirements such as ambient temperature.
Date When Full Compliance Will be Achieved:
PBNP was in compliance for these apparent violations when the aforementioned procedure
changes were completed in July,1997.
Apparent Violation 3:
The follow'mg is an apparent violation of Appendix B, Criterion V requirements that procedures
contain appropriate qualitative acceptance criteria for determining that important activities have
been satisfactorily accomplished. The licensee failed to provide adequate bus stripping procedure
l guidance as described in CR 96-959. Post-fire safe shutdown Abnormal Operating Procedures
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(AOPs)-10A and -10C directed the operators to the wrong DC power panel (D1l). Isolating the
wrong power supplies had the potential to cause r, urious operation of equipment powered from
buses 2A05 and 2B03 during a fire.
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Wisconsin Electric Response to Annarent Violation
We agree that the violation occurred as described. This violation was ider.tified by our staffin
Condition Report CR 96-959 and reported to the NRC in Licensee Event Peport (LER) 266/97-
020-00. This LER described the reason for the violation and the related corrective actions. I
Those reasons and corrective actions from the LER are summarized below and supplemented with
new information as necessary. Please note that a supplement to that LER will be issued on or
before October 14,1997.
Reason for Apparent Violation:
This procedural inadequacy was caused by failure to appropriately update a post-fire safe
shutdown procedure following modifications that installed new emergency diesel generators (G03
and G04). These design modifications changed the normal and alternate Vital DC power supplies
for the control power of safe shutdown 4KV switchgear 2A05 and 480 V bus 2B03. Without tha
procedure change, the operator was directed to the previously-used Vital DC power panel.
Corrective Actions Taken:
The appropriate post-fire safe shutdown procedure (AOP-10A," Safe Shutdown - Local
Control") was revised to properly direct operators to the appropriate Vital DC control panel for I
isolating control power to 4 KV bus 2A05 and 480 V bus 2B03. Note that the LER also
discussed the need to revise abnormal operating procedure AOP-10C because it too referred to
the wrong control panel. However, AOP-10C,"4160 Volt Vital Switchgear Room
inaccessibility" was t . 3d and the essential elements of the procedure were incorporated into
AOP-10A. In summary, AOP-10A has been revised and AOP-10C has been deleted such that
operators are no longer directed to the wrong DC panel. l
Corrective Actions to be Taken: l
The Appendix R Rebaselining Project will address the programmatic failures which caused the
apparent violations identified in the Inspection Report 1
Date When Full Compliance Will be Achieved:
PBNP was in compliance for this apparent violation when the aforemertioned procedure change
was completed in July,1997.
Apparent Violation 4:
! Emergency lighting was not provided in the Circulating Water Pump House in the areas where
operators had to manually start the diesel fire pump and to manually align the service water
system to provide cooling water to the turbine-driven AFW pump bearings. The failure to
identify that emergency lighting units were required in the Circulating Water Pump House is an
apparent violation of Appendix R Section III.J which requires, in part, that emergency lighting
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units with at least an 8-hour battery power supply be provided in all areas needed for operation of
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Wisconsin Electric Response to Apparent Violation
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We agree that the violation occurred as described. This violation was identified by our staffin
Condition Repon 97-1509 and reponed to the NRC in Licensee Event Report (LER) 266/97-
023-00. This LER described the reason for the violation and the related corrective actions.
Those reasons and corrective actions from the LER are summarized below and supplemented with
new information as necessary.
Henson for Apparent Violation:
The failure to comply with 10 CFR 50 Appendix R Section Ill.J was caused when alternative
provisions were made in the original safe shutdown analysis without the appropriate regulatory
exemption. The original approach to illuminate access / egress and the interior lighting for the
manual actions prescribed for the Circulating Water Pump House relied on portable hand-held
lighting units that were made available to operators performing post-fire shutdown tasks. These
hand-held units have been administratively controlled and maintained near the main control room;
readily accessible for the operators who may be assigned to transit to the exterior buildings to
conduct manual actions. These hand-held units are specifically dedicated for operator use in
outside plant areas during plant fires, and are maintained appropriately. Notwithstanding these
provisions, Wisconsin Electric determined that an exemption to the Appendix R Section III.J
requirement should have been submitted.
Corrective Actions Taken:
During the summer of 1997, emergency lighting was installed in the circulating water
pumphause to ensure compliance with Appendix R Section III.J.
Corrective Actions to be Taken:
The Appendix R Rebaselining Project will address the programmatic failures which caused the
apparent violations identified in the Inspection Report.
Date When Full Compliance Will be Achieved:
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PBNP was in compliance for this apparent violation when the aforementioned procedure changes
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Generic Considerations
As described in LER 266/97-020-00, and as a result ofidentified weaknesses in our Fire
Protection Program including these apparent virlations, Wisconsin Electric has initiated an
Appendix R Rebaselining Project to address the general problems in the documentation and
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implementation of the existing Appendix R Safe Shutdown Analysis. This project was discussed
with NRC Region 3 Management on April 24,1997. This project will re-evaluate the current
design ofPoint Beach Nuclear Plant, Unit I and Unit 2 against the requirements of Sections III.G,
Ill.J, III.L, and III.O of 10 CFR 50, Appendix R. This re-evaluation will determine that the
ability to achieve and maintain safe shutdown conditions is ensured and that adequate
documentation and procedures are provided to demonstrate the safe shutdown capability. It is
expected that the Rebaselining Project will provide programmatic improvements in the areas of
configuration management, modification control, and training. These program improvements
should address the root causes of the conditions reported herein. If determined necessary,
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additional design modifications will be implemented. We expect that the Appendix R
Rebaselining Project will be completed in June of 1999. The detailed schedule for this plan was
provided to the NRC in our letter dated October 10,1997.
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Y POWER COMPANY
Point Beach Nucear Plant (920) 755-2321
6610 Nuclear Rd.. Two Rivers, WI 54241
NPL 98-0098
February 11,1998
Document Control Desk
U. S. NUCLEAR REGULATORY COMMISSION !
Mail Station PI-137
Washington, DC 20555
Ladies / Gentlemen: I
DOCKETS 50-266 AND 50-301
REPLY TO A NOTICE OF VIOLATION
NRC INSPECTION REPORTS 50-266/97010 AND 50-301/97010
POINT BEACII NUCLEAR PLANT. UNITS 1 AND 2
In a letter from Mr. John A. Grobe dated September 15,1997, the Nuclear Regulatory j
Commission forwarded the results of an inspection conducted by your staff at our Point Beach i
Nuclear Plant from May 19,1997, through June 13,1997. This inspection report included a
Notice of Violation that identified two violations of NRC requirements.
In our response dated October 1 ~ ,1997, we did not agree that Violation 2 was a violation of i
NRC requirements. Violation 2 stated that a welding procedure specification used during our
Unit 2 Steam Generator Replacement Project did not meet the provisions of QW-409.1 as
delineated in ASME Section IX. Violation 2 noted that the welding procedure specification
allowed welding to be performed with an increase in heat input over that which had been
demonstrated through the Charpy impact testing of the qualification weld. Our October 15,
1997, response summarized recent ASME Main Committee and Section IX Subcommittee
reviews and actions regarding QW-409.1 provisions, our interpretation of applicable Code
requirements, and our basis for disputing the Violation 2.
On November 4,1997, members of our staff participated in a conference call with NRC
Region III personnel to further discuss Violation 2. This additional dialogue provided us a better
understanding of the Nuclear Regulatory Commission's technical position and philosophy
regarding the intent of QW-409.1. We continue to believe that a prevalent industry interpretation
or broad consensus regarding the code compliant application of QW-409.1 does not exist.
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llowever, we understand the Commission's technical position and acknowledge that it represents
a fundamentally sound interpretation, and accordingly, we accept the violation. We now agree
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j that the example cited is a violation of NRC requirements as contained in 10 CFR 50, Appendix
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B, Criterion IX. Accordingly we are submitting the attached revised response to Violation 2.
FEB17 1SSS
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L If you have any questions or require additional i-formation regarding this response, please
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Sincerely,
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cott A. Patuiski
Site Vice President
Attachment
cc: . NRC Regional Administrator
l NRC Resident Inspector
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NRC Project Manager
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DOCKETS 50-266 AND 50-301
REPLY TO A NOTICE OF VIOLATION
NRC INSPECTION REPORTS 50-266/97010 AND 50-301/97010
POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2
During an inspection conducted from May 19,1997, through June 13,1997, two violations of
NRC requirements were identified. Inspection Reports 50-266/97010 and 50-301/97010 and the
Notice of Violation transmitted to Wisconsin Electric on September 15,1997, provide details
regarding the violations. Wisconsin Electric submitted a response to the inspection report and
enclosed Notice of Violation via a letter to the Nuclear Regulatory Commission dated October
15,1997. In our response, we did not agree Violation 2 was an example of a violation of NRC
requirements as contained in 10 CFR 50, Appendix B, Criterion IX.
. Upon further review and as a result of additional dialogue between representatives of the Point
Beach Nuclear Plant staff and the NRC inspectors, we now agree that Violation 2 is a violation
of NRC requirements.
Accordingly, this revised response addresses Violation 2. In accordance with the instructions
provided in the Notice, our reply to the alleged violation includes: (1) the reason for the
violation, or if contested, the basis for disputing the violation; (2) corrective action taken; (3)
correctiu action to be taken to avoid further violations; and (4) the date when full compliance .
will be achieved.
Violation 2
"10 CFR 50, Appendix B, Criterion IX, " Control of Special Processes," requires, iu part, that
measures shall be established to assure that special processes, including welding, are controlled
and accomplished by qualified personnel using qualified procedures in accordance with
applicable codes, and standards.
ASME Section IX,1995 Edition, requirement QW-409.1, as implemented by requirement
QW-256, requires that "A change in the type of current or polarity, an increase in heat input, or
an increase in volume of weld metal deposited per unit length of weld, [is not allowed] over that
qualified."
Contrary to the above, the inspectors identified on June 13,1997, that weld procedure l
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WPS GT-SM/3.3-2 PB, Revision 1, did not meet the requirements of QW-409.1. The weld
procedure allowed welding to be performed with an increase in heat input over that which had
been demonstrated through Charpy impact testing of the qualification weld. l
This is a Severity Level IV violation (dupplement 1)."
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Response to Violation 2
We concur that Violation 2 is a violation of NRC requirements as characterized in the inspection
report. The circumstances surrounding the event are as stated in Inspection
Reports 50-266/97010 and 50-301/97010 and in our initial response to the Notice Of Violation
dated October 15,1997.
Reason for violation
Wisconsin Electric and our Steam Generator Replacement Project construction contractor
believed that welding procedure specification WPS GT-SM/3.3-2 PS and associated procedure
qualification record PQR GT-SM/3.3-Q2 conformed to the requirements of the 1995 edition of
ASME Section IX and the 1986 edition of ASME Section III, Subsection NB, regarding
establishing the heat input as required by QW-409.1. The welding procedure specification and
associated procedure qualification record were formulated and applied during our Unit 2 Steam
Generator Replacement Project in what was thought to be in accordance with a common
interpretation of the provisions QW-409.1 and accepted industry practices.
Recent reviews of QW-409.1 by the ASME Main Committee and the ASME Section IX
Subcommittee demonstrate that a broad consensus regarding the application of the provisions of
QW-409.1 has not veen reached. It is apparent that technical experts both within and outside the
ASME organization are not in agreement on the requirements for, or the practical
implementation of, QW-409.1. We believe that the absence of consensus betwen industry,
ASME, and NRC contributed to our nonconformance with the Code interpretation nade by the
Commission in support of this violation.
Addit' i onal background information regarding the recent reviews and actions of the ASME Main
Committee and ASME Section IX Subcommittee are pertinent to the violation. Our
understanding of these reviews and actions are summarized in the following paragraphs.
The concern raised by the NRC is similar to that raised by the author of Code
Interpretation IX-92-69. This interpretation was initially received by ASME in January,1992,
and various parts were answered over the next several Code meetings. The main issue of
controversy that remained ur resolved regarded the location ofimpact test specimens relative to
the various heat inputs that might occur at various locations in a test coupon joint.
In the May,1993, meeting, the inquiry was approved by the Section IX Subconunittee, subject to
the approval by the Main Committee. At the same meeting, proposed changes to ASME
Section IX, QW-409.1 were also approved by the Secuon IX Subcommittee. These changes
would have changed the article to read as follows:
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l "A change in the type of current or polarity, an increase in heat input, or an increase in
volume of weld metal deposited per unit length of weld, over that qualified. The
maximum heat input or volume of weld metal per unit length of weld qualified shall be
determined from the weld passes with the nominal heat input or volume of weld metal per
unit length or weld sampled by the impact test specimens."
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Note that this wording is consistent with the concern raised by the NRC. It should be noted,
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though, that this interpretation generated significant controversy within both the ASME
Section IX Subcommittee and the Main Committee. Although the interpretation and the
proposed change were passed by the Section IX Subcommittee, three negative votes were noted,
based primarily on the practical implementation of the proposed changes. Furthermore, the Main
Committee registered fifteen negative votes, and both the interpretation and the proposed
changes to QW-409.1 were voted down at the Main Committee level.
Because the interpretation was voted down at the Main Committee level, the interpretation
should not have been published. Many older interpretations were not well received by the main
Committee and should not be accepted carte blanche. Shortly after this interpretation was
published, ASME changed their rules of operation such that interpretations are not published
until the Main Committee has approved both the interpretation and any corresponding changes to
the Code. Efforts are currently underway to review outstanding interpretations, and either
incorporate or withdraw the interpretations. In fact, Interpretation IX-92-69 was reviewed during
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a September 16,1997, meeting of the ASME Section IX Subcommittee. It is our understanding
that the Section IX Subcommittee voted unanimously to withdraw the inquiry.
Wisconsin Electric consulted with two acknowledged ASME Code experts with the intent to
facilitate closure of the issue. Both consultants arrived at the same conclusion; namely, that the
subject procedure qualification record as originally performed, was in full compliance with the
ASME Boiler and Pressure Vessel Code. Further, one of the consultants generated the request
for additional review and subsequent recent withdrawal of Code Interpretation IX-92-69 by the
ASME Section IX Subcommittee.
While we understand the Nuclear Regulatory Commission's technical position and acknowledge
that the Commission's interpretation is fundamentally sound, the above summary demonstrates l
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that a widely held understanding and consensus on this issue has not been achieved.
Corrective Actions Taken
The subject issue associated with WPS GT-SM/3.3-2 PB was initially identified by the NRC as
an unresolved item in Inspection Report 50-301/96014 (DRS) dated February 7,1997. It was
noted in Inspection Report 50-301/96014 (DRS) that based upon discussions with the Office of
Nuclear Reactor Regulation and the Office of Nuclear Regulatory Research technical staff and ;
reviews of NDE results, the inspectors did not have a concern for the technical adequacy of the )
girth welds. A conference call with members of our staff and NRC representatives was ;
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conducted on May 7,1997, to further review the issue and to discuss Code compliance j
interpretations. Although this conference call did not fully resolve the issue, Wisconsin Electnc
decided to take immediate conservative actions to assure that the Code compliant status of welds
made by WPS GT-SM/3.3-2 PB would not be in doubt prior to the restart of Unit 2.
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At the request of Wisconsin Electric, our Steam Generator Replacement Project construction
contractor performed another qualification weld and additional associated Charpy impact testing.
This additional qualification weld was performed to requalify the gas tungsten arc weld (GTAW)
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portion of the qualification record. The additional qualification weld is documented as PQR
GT/3.3-Q2 dated May 20,1997. The requalification weld coupon used a heat input range that
included the maximum listed on WPS GT-SM/3.3-2 PB. These maximum heat inputs were
within the area where the impact testing coupon was taken to perform qualification testing of
PQR GT/3.3-Q2.
Welding procedure qualification testing for PQR GT/3.3-Q2 was conducted in accordance with
applicable ASME Section III and Section IX requirements, including Charpy impact testing. The
results of the qualification testing for PQR GT/3.3-Q2 are documented in a test report dated May
27,1997. The successful requalification satisfactorily demonstrated the code qualification of
weld procedure WPS GT-SM/3.3-2 PB. Although performed by our construction contractor, this
additional qualification weld and associated testing was performed under the scrutiny of a
Wisconsin Electric Quality Assurance representative.
Wisconsin Electric directed our construction contractor to review all other impact tested
procedure qualification records and related site specific welding procedure specifications
prepared for our Unit 2 Steam Generator Replacement Project to assure that no other weld
qualifications were at issue. Our contractor confirmed, in a letter dated June 6,1997, that the
violation issue did not apply to all other impact tested procedure qualification records which
were used for permanently installed welds. For these other qualification records, the contractor
confirmed that the highest heat input listed on the procedure qualification records were within the
area where the Charpy impact testing coupon would have been taken and that heat input
allowables for associated welding procedure specifications were at or below that allowed by the
procedure qualification record.
Members of our engineering, maintenance and quality assurance stafTthat deal with welding i
issues and/or qualification have been notified of the issues surrounding this violation.
Date of Full Compliance I
Full compliance with NRC requirements was achieved on May 27,1997, upon completion of
qualification weld PQR GT/3.3-Q2 and associated welding procedure qualification testing which
satisfactorily demonstrated the code qualification of WPS GT-SM/3.3-2 PB. No additional
actions are required to achieve full compliance with NRC requirements.
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