ML20113F494
| ML20113F494 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 09/19/1996 |
| From: | Hill W NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9609240262 | |
| Download: ML20113F494 (11) | |
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Northem States Power Company Monticello Nuclear Generating Plant 2807 West Hwy 75 Monticello, Minnesota 55362 9637 September 19,1996 10 CFR Part 2 Section 2.201 US Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Reply to Notice of Violation Contained in NRC Inspection Report No. 50-263/96006 Pursuani to the provisions of 10 CFR Part 2, Section 2.201, our reply to the notice of violation contained in your letter of August 20,1996, is provided in Attachment A. In addition, the August 20, 1996 letter contained a notice of deviation from licensee commitments. Our reply to the notice of deviation is provided in Attachment B.
Attachment A, Reply to Notice of Violation, contains the following new NRC commitments:
A review will be performed of Operations Manual, Volume B, procedures to identify procedures similar to B.4.3.2-05.G which are relied upon to establish post accident )
equipment configurations necessary to mitigate the consequences of design basis accidents. i A further review will be conducted of these procedures to confirm that similar procedure inadequacies do not exist. This action is to be completed by November 30,1996.
In addition, Operations Manual, Volume B, procedures which are identified as relied upon to establish post accident equipment configurations necessary to mitigate the consequences of design basis accidents are to be further reviewed by the Operations Training Advisory Committee to determine if additional training needs are required for these procedures. This action is to be completed by January 31,1997.
Procedures are to be revised to include testing of the logic circuit bypass of the CGCS booster pump motor and CGCS strainer motor thermal overloads. This action is to be completed by January 31,1997.
Attachment 8, Reply to Notice of Deviation, contains the following new NRC commitments:
l The Monticello ISI Plan is to be revised, as appropriate, to reflect the performance of an l augmented examination of the reactor vessel welds in accordance with 10 CFR l 50.55a(g)(6)(ii)(A), to the maximum extent practical, without imposing additional hardship or burden on NSP without a compensating increase in the level of quality or safety for those l examinations which may be deemed to be impractical. The ISI plan will be revised prior to the next Monticello refueling outage (January 1998).
D/19/96 JMM XT04560WloNTILMi\VloL\96006R1. Doc <b DJ lf I 9609240262 960919 PDR ADOCK 05000263-G PDR
l USNRC NORTHERN STATES POWER COMPANY September 19,1996 '
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NSP will submit a relief request in lieu of invoking IWA-2240 for procedure NSP-UT-16. The relief request will be submitted by November 15,1996.
Please contact Marvin Engen, Sr Licensing Engineer, (612-295-1291) if you require further ,
information.
i Nf4 fh 4 William J Hill l Plant Manager Monticello Nuclear Generating Plant c: Regional Administrator- 111, NRC NRR Project Manager, NRC l Sr Resident inspector, NRC State of Minnesota Attn: Kris Sanda J Silberg Attachments A - Reply to Notice of Violation B - Reply to Notice of Deviation c
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Attachment A !
REPLY TO NOTICE OF VIOLATION j Violation i Technical Specifications (TS) Section 6.5, " Plant Operating Procedures," required that i detailed written procedures covedng plant operations areas shall be prepared and followed. '
i TS Section 6.5.A.3 required detailed written procedures covering actions to be taken to ,
correct specific and foreseen potential malfunction of systems or components, including i follow-up actions required afterplant protective system actions have initiated.
TS'Section 6.5.A.4 required detailed written procedures covering surveillance and testing requirements that could have an effect on nuclear safety.
I Contrary to the above:
- a. Operations ManualProcedure B.4.3.2-05.G, " Combustible Gas Control System l (CGCS), Post-LOCA CGCS Startup and Operation," Revision 4, did not provide )
adequate instructions formanual operatorinitiation of the CGCS system i following a loss of coolant accident. The operatorinstmctions would have prevented the inlet and outlet isolation valves from opening on demand. l
0417-1(2)AB CGCS Recombiner Tests, Revision 3, were inadequate in that :
neither required the simulation of a primary containment Group ll Isolation signal during testing to demonstrate that the isolation signal would be automatically bypassed upon system demand as designed.
l NSP Response to Example A NSP acknowledges example 'a' cited in the above Notice of Violation.
Reason For Violation:
The reason Operations Manual Procedure B.4.3.2-05.G, " Combustible Gas Control System !
(CGCS), Post-LOCA CGCS Startup and Operation," did not provide adequate instructions, is that information which was available during preparation of the procedure was not used correctly during procedure creation in 1984.
The system specification provided in specification NSP-63-130, which was available and ,
referenced in the revision 0 of the CGCS Operations Manual, correctly described the switch ,
manipulations for opening the CGCS containment isolation valves. In addition, the elementary logic diagrams correctly shows that, with a Group 11 containment isolation signal l
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September 19,1996 present, both the inboard and outboard containment isolation valve position switches must be
- in the open position prior to taking the control switches to the bypass position to establish l proper relay logic.
The control switch for the CGCS containment isolation valves is a three position switch with the positions CLOSED, OPEN, AND BYPASS. The inboard and outboard containment :
isolation valves are controlled by separate control switches. The preparer of the CGCS !
Operations Manual startup procedure may have thought that taking each control switch from i CLOSED through OPEN to BYPASS was sufficient to initiate the logic, or confusion may have l been introduced from the use of the pre-operation test procedure as a basis for establishing the CGCS startup procedure. The pre-operation test, tested the containment isolation i function of the CGCS inboard and outboard containment isolation valves by establishing the valves in the open position, inserting a simulated Group 11 containment isolation signal, i l Verifying the valves closed, and then directed to place the control switch for the containment 1 l Isolation valves in the bypass position with a verification that the valves opened. The l information from the pre-operation test procedure may have been misinterpreted in that it was j not recognized that the control switches for the CGCS inboard and outboard containment i isolation valves had been placed in the open position in order to obtain the desired system L response with the Group 11 containment isolation signal present.
Corrective Action Taken and Results Achieved: ;
A temporary procedure change was implemented on June 7,1996 to provide the correct '
procedural controls in Operations Manual Procedure B.4.3.2-05.G. " Combustible Gas Control ;
i System (CGCS), Post-LOCA CGCS Startup and Operation," for opening the CGCS system l containment isolation valves for initiation of the CGCS system following a postulated loss of ;
coolant accident. Operations Manual Procedure B.4.3.2-05.G was revised to incorporate this temporary procedure change with Revision 5, approved on August 23,1996. l L
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Corrective Action to be Taken to Avoid Further Violations -
A review will be performed of Operations Manual, Volume B, procedures to identify procedures similar to B.4.3.2-05.G which are relied upon to establish post accident equipment configurations necessary to mitigate the consequences of design basis accidents. A further review will be conducted of these procedures to confirm that similar procedure inadequacies l do not exist. This action is to be completed by November 30,1996.
Operations Manual, Volume B, procedures which are identified as relied upon to establish ,
post accident equipment configurations necessary to mitigate the consequences of design basis accidents are to be further reviewed by the Operations Training Advisory Committee.
This review will be performed to determine if additional training needs are required for these procedures. This action is to be completed by January 31,1997.
Plant administrative controls goveming procedure review and approval were reviewed for adequacy. It was confirmed that adequate guidance is contained in the current administrative
- controls. Since the time of initial creation of Operations Manual procedure B.4.3.2-05.G, l guidance has been added to the administrative controls regarding validation of procedures,
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Attachment A Page 3 September 19,1996 including validation on the plant simulator when feasible, as well as guidance on the performance of reviews for required training on new or revised procedures.
Date When Full Compliance Will Be Achieved Full compliance has been achieved.
NSP Response to Example B Conceming example 'b' cited in the above Notice of Violation, we concur that surveillance test procedures 0255-21111-1(2), A(B) CGCS Tests, and 0417-1(2), A(B) CGCS Recombiner Tests, do not require the simulation of a primary containment Group 11 isolation signal during :
testing. However, Monticello surveillance test procedure 0150, " Group 2 & SCTMT isolation l Simulated Automatic initiation Test," does contain procedure steps which demonstrate the CGCS design feature which allows bypassing the primary containment Group 11 isolation l signal during CGCS system initiation. Procedure 0150 was revised via revision 19, approved ;
l May 8,1996, to require the simulation of a primary containment Group 11 isolation signal ,
i during testing to demonstrate that the isolation signal would be automatically bypassed upon j j system demand as designed. Procedure 0150 is performed each refueling outage. The procedure satisfactory tested the CGCS bypass of the primary containment Group 11 isolation signal during the 1996 refueling outage.
l The CGCS is designed to allow operation of the system with a Group 11 primary containment j isolation signal following a postulated Loss of Coolant Accident (LOCA). The controllogic circuit of the CGCS provides for bypassing of the Group 11 primary containment isolation signal to allow opening of the CGCS containment isolation valves. In addition the logic provides for bypassing the protective function of the thermal overloads for the CGCS cooling water booster pump and the automatic basket strainer. The thermal overloads are safety l grade devices. The Monticello plant staff self identified the procedure enhancement to provide testing of the relays and contacts associated with bypassing the Group !! primary containment isolation signal. However, during further review of the issue identified in example
'b' of the above Notice of Violation, the Monticello plant staff identified that these procedure ,
enhancements did not include testing to demonstrate the bypassing of the thermal overloads !
for the CGCS cooling water booster pump and the automatic basket strainer. Wnile NSP disputes the specific details cited in example 'b', we acknowledge the intent of example 'b'
!. cited in the above Notice of Violation.
I Reason For Violation:
l Plant procedures do not include steps to verify that the protective function of the safety grade thermal overloads for the CGCS cooling water booster pump and the automatic basket strainer are bypassed; however, these contacts are actuated from the same relay tested per procedure 0150 to demonstrate bypassing of the primary containment Group !! isolation signal
- during CGCS system initiation. The safety grade thermal overloads are tested in accordance
! with the Montice!!o preventative maintenance program to ensure proper operation to support CGCS operation following a postulated LOCA. Because the bypass logic for the thermal I
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l , Attachmint A i Page 4 September 19,1996 overloads is redundant to the safety grade thermal overloads, testing of this logic was not 3
cons;dered to be necessary.
I j Corrective Action Taken and Results Achieved: i
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Monticello has performed testing which confirmed that the bypass of the thermal overloads i
- will function as required. This testing confirmed the functionality of the parallel bypass path in i the CGCS cooling water booster pump and the automatic basket strainer initiation logic circuit, I
which is redundant to the safety grade thermal overload devices.
Corrective Action to be Taken to Avoid Further Violations:
! The Monticello plant staff is aware of NRC concems identified in NRC Generic Letter 96-01, l
, " Testing of Safety-Related Logic Circuits," regarding ensuring adequate surveillance
, procedures for the testing of logic circuitry in accordance with plant Technical Specification requirements, and will address the concems identified in Generic Letter 96-01 as stated in our j response to the generic letter.
Procedures are to be revised to include testing of the logic circuit bypass of the CGCS 4 booster pump motor and CGCS strainer motor thermal overloads. This action is to be j completed by January 31,1997.
Date When Full Compliance Will Be Achieved Full compliance has been achieved.
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l Additional Information Concernina issues Related to the Notice of Violat'on i f ,
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! Monticello Integrated Inspection Report 96006 identified two conditions which would render l CGCS inoperable. Those being the inadequacy of Operation Manual procedure B.4.3.2-05 l and the CGCS booster pump motor thermal overload concem. Monticello provides the following information to clarify CGCS operation during initiation of the system following a i postulated LOCA. The power cabinets for the CGCS system are located in the Reactor Building and contain equipment to supply power for operation of the recombiner unit and '
recombiner control panel. Two redundant power supply cabinets are provided for the i redundant CGCS divisions. Contained within the power cabinet are the CGCS heater i
! controller, blower and motor starters, and the power supply for the recombiner control panel.
[ The power cabinets are de-energized whenever the recombiner is not functioning. The power cabinet for a division of CGCS only becomes energized when the control switches for the containment isolation valves associated with that division of CGCS are in the OPEN position.
I Monticello recognizes that the lack of adequate procedural guidance in Operation Manual Procedure B.4.3.2-05.G could result in a delay in initiation of the CGCS system; however we
, do not consider the performance trend of the thermal overload devices for the CGCS booster pump to create an additional concem for CGCS system initiation. During conditions when the
- CGCS system is initiated for postulated accidents, the CGCS system will not have power supplied to the system and thus will not operate unless control switches for the containment i
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Attachmtnt A l Page 5 l September 19,1996 isolation valves are in the open position. Once the logic is properly established to open the containment isolation valves, the power supply cabinet is energized and the thermal overload !
I device for the CGCS booster pump is bypassed.
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Attachment B REPLY TO NOTICE OF DEVIATION Deviation A. In an NRC letterdated October 18,1994, the NRC accepted Northem States Power l commitments to follow ASME Code Section XI,1986 Edition requirements for the third ten-yearintervalinservice inspection (ISI) program. ForISIplans, paragraph IWA-2420 of Section XI,1986 Edition of the ASME Code specified that each inspection plan shallinclude the Edition and Addenda of this Division that apply to the required examinations and tests.
Contrary to the above, on June 6,1996, the licensee's ISIplan did not include the ASME Section XI,1989 Edition Table IWB-2500-1 requirements applicable to reactor vesselshellwelds.
NSP Response NSP acknowledges the above deviation.
Reason for Deviation The 1986 Edition of the ASME Code requires that a percentage of the reactor vessel shell welds be inspected. This percentage is scheduled. Following development of the Third Ten Year Inservice Inspection Plan, the NRC issued the requirements for augmented reactor vessel examinations,10 CFR 50.55a(g)(6)(ii)(A), which requires 100% of the reactor vessel shellwelds be examined to the 1989 Edition of the ASME Code by the end of the existing interval. The deviation states that the augmented inspections have not been scheduled, nor has the 1989 Edition of the ASME Code been referenced. The following is provided in response to the deviation.
(a) The Monticello third ten-year interval inservice inspection (ISI) program is in compliance with ASME Section XI,1986 Edition (53FR16501). Revision 0 of the Monticello ISI program was submitted by NSP letter dated December 31, 1991. Revision 1 of the program was submitted to the NRC for review by NSP letter dated July 29,1993 with additionalinformation provided by NSP in our letter dated December 20,1993. Monticello's submittal included a request for relief from the Code requirements of IWB-2500-1, which requires a 100%
volumetric examination of one circumferential and one longitudinal beltline region weld.
- (b) The NRC evaluation of the Monticello ISI program plan for the third 10-year l
Interval, provided by NRC letter dated October 18,1994, accepted the
! Monticello ISI plan. However the NRC evaluation denied the request for relief conceming requirements of IWB-2500-1, based on the augmented reactor pressure vessel examination requirements of 10 CFR 50.55a(g)(6)(ii)(A), which was effective on September 8,1992 and invoked the 1989 Edition of the Code for the augmented inspection.
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(c) Although the Monticello ISI Plan was submitted and accepted by the NRC staff ,
without specific identification of the requirements in 10 CFR 50.55a(g)(6)(ii)(A), I
" Augmented Examination of Reactor Vessel", we are aware of the augmented reactor vessel examination requirements as noticed in Federa/ Register ,
Volurne 57,34666 and the NRC Monticello ISI program evaluation. NSP has '
representatives on the BWROG Committees which are currently working on the 1 examination technique to implement the augmented reactor vessel inspection l activities.
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(d) ASME Section XI,1986 and 1989 Editions per Table IWB-2500-1 category B-A, reactor vessel shell welds, as well as 10 CFR 50.55a(g)(6)(ii)(A), allows for a deferral of the inspection to end of the interval, which ends May 31,2002.
Corrective Action Taken and Results Achieved Northem States Power Company (NSP) has representatives on the BWROG Committees which are currently working on the examination technique to implement the augmented reactor vessel inspection activities.
l Corrective Action to be Taken to Avoid Further Deviations )
The deviation indicated that not all of the reactor vessel shell welds were identified for examination in the ISI plan, nor was the Code to which the augmented inspections to be i examined referenced. The Monticello ISI Plan is to be revised, as appropriate, to reflect the performance of an augmented examination of the reactor vessel welds in accordance with 10 CFR 50.55a(g)(6)(ii)(A), to the maximum extent practical, without imposing additional hardship i or burden on NSP without a compensating increase in the level of quality or safety for those !
examinations which may be deemed to be impractical. I Date When Corrective Action will be Completed The ISI plan will be revised prior to the next Monticello refueling outage (January 1998).
Deviation B. In an NRC letter dated October 18th,1994, the NRC accepted Northem States Power commitments to follow ASME Code Section XI,1986 tidition requirements for the third ten-yearintervalinservice inspection program. For ultrasonic (UT) examinations of dissimilar metal welds, paragraph Ill-3411(b) of Appendix Ill of Section XI,1986 1 Edition of the ASME Code specified the following: "If the examination will be conducted from both sides, calibration reflectors shall be provided in both materials."
Contrary to the above, on June 6,1996, procedure ISI-UT-16, "UT Examination of Welds of Austenitic and High Nickel Alloy Materials," Revision 11, specified that UT examinations conducted from both sides of dissimilarmetal welds could be calibrated utilizing reflectors in a single calibration block material. As a result, several UT
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September 19,1996 !
l examinations were completed on dissimilarmetal welds utilizing a single calibration block material.
1 NSP Response j NSP acknowledges that additional calibration blocks could have been procured or a relief request submitted to address the Appendix Ill requirement. However, NSP believes we are in full compliance with our ISI Program commitments through the proper use of IWA-2240.
The use of a single stainless steel block for an examination performed from both sides of a stainless to carbon steel dissimilar weldment was implemented by invoking Section XI article
' IWA-2240, which states:
"Altemative examination methods, a combination of methods, or newly developed techniques may be substituted for the methods specified in this Division, provided the Inspector is satisfied that the results are demonstrated to be equivalent or superior to those of the specified method."
In accordance with IWA-2240, the implementing procedure, ISI-UT-16, "UT Examination of Welds in Austenitic and High Nickel Alloy Materials," was demonstrated to the Authorized . l Inspection Agency which agreed the examination results were equivalent or superior to the !
Appendix 111 requirements. The procedure was reviewed and approved by the anils and !
subsequent revisions were reviewed and approved by the ANil prior to use. The application of IWA-2240 was clearly documented and was observed by the NRC Auditor because it was flagged in the ISI Field Supervisor's records. ,
The application of IWA-2240 to Section XI appendices is supported by Code interpretation XI-80-08. A review of this article from the 1977 edition through the 1986 edition indicates that it has not changed. This use of a stainless steel calibration block for an examination from both ;
sides of a stainless / carbon steel dissimilar weld is considered prudent for the following reasons: l a) The sensitivity established on the stainless block is equal to or higher than that :
which would have been obtained on a carbon steel block providing a conservative examination sensitivity, b) The velocity difference between stainioss and carbon is minor and does not result in an appreciable angle variation.
c) The procedure provides for reexarnination using a calibration block of similar material if indications are noted that are not attributable to geometry.
d). Use of two separate calibrations may result in greater radiation exposure to personnel involved in the examination.
e) The Appendix lit requirement to use two calibration blocks was new with the implementation of the third ten year interval ISI program in June 1992. The use
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,. Att: chm:nt B Page 4 September 19,1996 of less sensitive calibration blocks for dissimilar weldments was considered to be less effective.
All ISI program examinations that invoke IWA-2240 were verified to meet the requirements of that article. NSP feels that we are in full compliance with our previously stated commitments; i however, NSP will submit a relief request in lieu of invoking IWA-2240 for procedure NSP-UT- l
- 16. The relief request will identify all applicable weldments. Until a response to this relief request is received, we will continue to implement the existing procedure for the reasons !
stated above. The relief request will be submitted by November 15,1996.
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