IR 05000245/1998208
| ML20237B076 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/12/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20237B068 | List: |
| References | |
| 50-245-98-208, 50-336-98-208, 50-423-98-208, NUDOCS 9808180033 | |
| Download: ML20237B076 (108) | |
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U.S. NUCLEAR REGULATORY COMMISSION 1 REGION I f t
Docket Nos.: 50-245 50 336 50-423 Report Nos.: 98-208 98-208 98-208 License Nos.: DPR-21 DPR-65 NPF-49 Licensee: Northeast Nuclear Energy Company P. O. Box 128 Waterford, CT 06385 i
Facility: Millstone Nuclear Power Station, Units 1,2, and 3 Inspection at: Waterford, CT Dates: April 28,1999 - June 29,1998 Inspectors: T. A. Eas!ick, Senior Resident inspector Unit 1 D. P. Beaulieu, Senior Resident inspector, Unit 2 A. C. Cerne, Senior Resident inspector, Unit 3 P. Cataldo, Resident inspector, Unit 1 S. R. Jones, Resident inspector, Unit 2 B. E. Korona, Resident inspector, . Unit 3 E. B. King, Security inspector, Region i N. T. McNamara, EP inspector, Region l D. M. Silk, EP Inspector, Region 1 l
j L. J. Prividy, Senior Reactor Engineer, Region 1 l J. W. Andersen, Project Manager, NRR, HQ j P. P. Narbut, SPO, NRR, HQ j J. Brand, Seabrook Resident inspector L. L. Scholl, Senior Reactor Engineer, Reginn i l
N. J. Blurnberg, Project Engineer, Region i i P. R. Frechette, Security inspector, Region I j J. T. Furia, Senior Radiation Specialist, Region 1 {
D. G. Mcdonald, Project Manager, NRR, HQ j B. Hughes, NRR i T. G. Scarbrough, Mechanical Engineering Branch, NRR J. Higgins, NRC Contractor, BNL A. Fresco, NRC Contractor, BNL 1 P. Bezier, NRC Contractor, BNL I J. Cadwell, NRC Contractor, BNL
Approved by: Jacque P. Durr, Chief Office of the Regional Administrator Region I
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9008180033 900812 POR ADOCK 05000245'
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?< EXECUTIVE SU MMARY ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv u ,
J U1.1 Operations z. .
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[i' U101 Conduct of Operations' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 i' U108 Miscellaneous Operations issues (92700) . . . . . . . . . . . . . . . . . 3
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. U 1.Il Mein tenance r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 (m U1 M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 -
p U1 M2' Maintenance and Material Condition of Facilities and Equipment . . 5 L ,
' U1 M8 Miscellaneous Maintenance issues . . . . . . . . . . . . . . . . . . , . . . 7 .
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U 1.Ill Engineering . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
, U1 E3 Engineering Procedures and Documentation . . . . . . . . . . . . . . . . 8 L3 U1 E8 Miscellaneous Engineering lssues . . . . . . . . . . . . . . . . . . . . . . . 10
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' t U 2.1 Operations L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 L
U201 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 ,
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.U2 08 Miscellaneous Operations issues (92700) . . . . . . . . . . . . . . . . . .13 ;
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L , U 2.ll Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 - 1 U2 M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 l
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L U2.lli Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 .
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E U2 E1 . Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 - 1
- U2 E8 Miscellaneous, Engineering issues . . . . . . . .. . . . . . . . . . .. . . . . . 19 ' .
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, U3.1 Operations . .. ...............................................28 Conduct of Operations . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . 28
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U3 01-L U3 02 Operational Status of Facilities and Equipment ............35 '
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U3 03 Operations Procedures and Documentation . . . . . . . . . . . . . . . . 37 U3 07 Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . 38 1 7 q
, U3.H M ainte nan ce . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 6 l
' U3 M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . .' 46 U3 M3 ' Maintenance Procedures and Documentation . . . . . . . . . . . . . . 48 l
, U3 M8 Miscellaneous Maintenance issues . . . . . . . . . . . . . . . . . . . . . . 54 i U3.Ill Engineering .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 8 ;
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U3 E1 Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 l
. U3 E2 Engineering Support of Facilities and Equipment ............ 62 U3 E3 Engineering Procedures and Documentation . . . . . . . . . . . . . . . 71 U3 E7 . Quality Assurance in Engineering Activities . . . . . . . . . . . . . . . . 73
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- U3 ES Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . 75 i
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IV Plant Suppod .................................................82 R1 Radiological Protection and Chemistry Controls . . . . . . . . . . . . . 82 R5 Staff Training and Qualification in Radiological Protection and C h e mi s t ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 4 R7 Quality Assurance in Radiological Protection and Chemistry Activities
..............................................85 P8 Miscellaneous Emergency Preparedness issues . . . . . . . . . . . . . 85 S1 Conduct of Security and Safeguards Activities ............. 89 S2 Status of Security Facilities and Equipment . . . . . . . . . . . . . . . . 90 S3 Security and Safeguards Procedures and Documentation . . . . . . 91 S4 Security and Safeguards Staff Knowledge and Performance . . . . 91 S5 Security and Safeguards Staff Training and Qualification . . . . . . 92 S7 Quality Assurance in Security and Safeguards Activities ......93 V. M anagement Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4 X1 Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 94
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i EXECUTIVE SUMMARY Millstone Nuclear Power Station Combined Inspection 245/98-208:336/98-208:423/98-208 Operations e Blackness testing on the Unit 1 spent fuel pool storage racks to ensure the presence and integrity of the Boraflex panels was generally completed thoroughly and professionally. Conservative decision making and strong management oversight were apparent throughout the activities. All concerns raised by the inspectors were l immediately addressed. Daily planning meetings and pre-job briefings were noted ( strengths for this evolution. Health Physics support, including foreign material
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exclusion control was exceilent. (Section U1.01.1)
e The inspector reviewed the corrective actions associated with ACR M1-96-0564 j_ concerning a negative trend ,with regard to the control of contractors at Millstone.
l These corrective actions included new procedures, a new organization for contract i administration, and an improvement plan with long term actions that are scheduled ( for completion at the erid of 1998. SIL item 39, which addresses this ACR, will
! remain open pending completion of the improvement plan and a review of the l effectiveness of the corrective actions. (Section U1.08.1)
i e The inspector reviewed the licensee's root cause evaluations and corrective actions L for the operational events documented in the Unit 3 Operational Safety Team
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Inspection, NRC Inspection Report 50-423/97-83, dated June 12,1998. The corrective actions were comprehensive and appropriately addressed the deficiencies identified in the root causo evaluations, as well as the operator performance issue In addition, the licensee's corrective actions for an OSTI concern related to the valve and system alignment program was reviewed. The licensee demonstrated i that the valve and system alignment program was effectively implemente ]
(Sections U3.01.2 and 01.3)
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l e Although the licensee's performance during several Unit 3 plant heatups and cooldowns was generally acceptable, the licensee violated their plant heatup procedure during one plant heat up and transition into Mode 4. (Section U3.01.4)
i Maintenance e The licensee's performance of the Unit 1 diesel generator surveillance was goo However, the generic implications of configuration management and human performance issues due to the #3 cylinder indicator valve operation without 1 procedural or operations department direction is considered unresolved pending L completion of the licensee's evaluation of this event. (Section U1.M1.1)
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- The inspector reviewed the licensee's follow up and corrective actions for a concern associated with an inconsistency between procedure RP-4, " Corrective Action ,
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Program" and WC-8, ' Control and Calibration of Measuring and Test Equipment" {
regarding the timeliness for evaluating M&TE NCRs which may result in a condition I adverse to quality. The corrective actions were adequate to resolve this issu (Section U1.M2.1)
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The Unit 3 material, equipment, and parts lists (MEPL) program was reviewed in severalinspections over the past year. The licensee has invested substantial effort into improving the program and has significantly upgraded both the program and the evaluations for many components and parts in Unit 3. A number of issues were identified during the review and the licensee has been responsive in addressing the concerns. The program is currently deemed acceptable and meets regulatory requirements. (Section U3.M3.1)
Englasering
While violations of NRC requirements have occurred relative to commercial grade dedication and receipt inspection activities of emergency diesel generator (EDG) fuel oil across all three Millstone units, no action is required pending the licensee's response to the similar violation contained within NRC Inspection Report 50-245/98-207. In addition, the inspector concluded the evaluations performed by the licensee concerning the applicability of the lube oil / fuel oil incompatibility on the EDGs at all three Millstone units was adequate to support tneir conclusions. (Section U1.E3.1)
- The licensee identified concerns with the review process for temporary modifications implemented through approved plant procedures. They issued a comprehensive corrective action plan, with one exception, to address future procedure changes that include temporary modifications. At Millstone Unit 2, the licensee failed to effectively address concerns with existing temporary modifications implemented through approved plant procedures. Specifically, the licensee's corrective actions for existing modifications was to make corrections during the !
biennial procedure review process. This did not appear to be timely for safety significant modifications. In the interim, the licensee issued a night order not to j implement existing modifications. This is an unresolved item. (Section U2.E1.3)
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- At Unit 2, the licensee's corrective actions were acceptable to address the concerns described in LER 50-336/96-33 which involved incorrect functional settings on the steam generator hydraulic snubbers. This condition could have resulted in exceeding design allowable stresses of the reactor cos lant system piping during postulated events. LER 50-336/96-33is closed. (Section U2.E8.1)
- At Unit 2, the licensee's corrective actions were acceptable in addressing a I technical specification non-compliance involving the automatic bypassing of the overload cutoffs on the refueling bridge for nine inches of vertical travel. Licensee performance was good in that this condition was identified by reviewing events at I
other facilities. This issue was characterized as a non-cited violation. (Section l U2.E8.2)
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l e At Unit 2, the licensee's corrective actions were acceptable to address the August l 2,1997, event in which the "A" EDG was rendered inoperable due to a vibration j induced weld joint failure that resulted in the spill of 5 to 7 gallons of tube oil before ;
the diesel was secured. The "A" and "B" EDGs were declared inoperable when it !
was discovered that the failed weld, as well as many other welds on the skid i mounted piping for both EDGs were found to be partial penetration welds (30% to 50% of the pipe wall thickness), not full penetration welds as delineated in vendor drawings. All welds in the large bore piping of both EDGs have been reworked to ;
full penetration welds bringing the diesels to a condition comparable to the condition j thought to exist at the time of initial purchase. Also, a vibration monitoring program was established to assure that exhibited vibrations meet the established limits. The NRC concluded that licensee's operability determination dated June 11,1998, that j considered the diesels to be operable and fully qualified was acceptable. (Section i U2.E8.3)
e Based on the review of the licensee's submittals, site audits, discussions with the !
licensee's staff, and the NRC staff's independent assessment of the current situation, the staff concluded that the erosion of cement from the underlying porous drainage system has not jeopardized the Unit 3 containment's ability to perform its safety function for the immediate future. Moreover, through an in-depth evaluation 2 of the present and future potential degradation of the porous concrete media, the I licensee demonstrated that the containment structure will maintain its ability to l perform the intended functions throughout the licensed life of the plant (until year l 2026), and beyond. (Section U3.E2.3) i I
- A review of the licensee's 10 CFR 50.59 safety evaluation process found that it is I adequate and should maintain effective configuration control of the Millstone 3 I licensing and design basis. (Section U3.E3.1) !
I e The licensee's corrective actions regarding the Unit 3 MOV thrust calculation violation were comprehensive. (Section U3.E8.6)
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I e At Unit 3, the RSS cubicle initial sump pump system design and the pump l qualification were inadequate. Significant corrective action, including system and j pump design changes, were required to ensure the system would perform its design i function. The inspectors also concluded that the immediate corrective actions taken i by the licensee were adequate to ensure system operability. (Section U3.E8.8) j l
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e The licensee established and implemented effective radiological protection programs l with respect to: (1) control and leak testing of sealed sources and maintaining an inventory of these sources; and, (2) maintaining decommissioning records required
, under 10 CFR 50.75(g) for areas located outside the protected are l l Decommissioning records for areas within the protected area are still under !
development. (Section R1) j l
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- Radiological controls were determined to be appropriate, especially in the areas of !
posting and control of high and locked high radiation areas, at all three unit Appropriate work planning for maintaining occupational exposures ALARA was also observed at Unit 3 for work on the 3RCS*V132 valve and for industrial radiography taking place in the containment. (Section R1)
- The licensee established an effective continuing training program for plant workers I and radiation protection technicians, inc!uding the use of a detailed mock-up facilit {
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Lesson plans and training objectives reviewed were appropriate with regards to I subject scope and depth of presentation. (Section RS)
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- The licensee has implemented an effective program for conducting annual audits and appraisals to meet the requirements of 10 CFR 20.1101(c). Additional self- I assessments are also performed by the various health physics organizations at the !
site. (Section R7) I i
e The licensee made significant improvements to the post accident sampling system (PASS). Procedures were rewritten, technicians retrained and the system was repeatedly tested. Also, as equipment deficiencies were identified, they were j corrected. With the exception of the dissolved gas sample results, the licensee met l the appropriate acceptance criterion. Corrective actions are sufficient to provide l reasonable assurance that the PASS system would be able to assist in the assessment of core damage, given a significant transient or accident. The licensee ,
is continuing to assess their method for retrieving and analyzing a dissolved gas !
sample. Until this issue is resolved, NRC Violation 50-423/98-01-01 will remain !
open. (Section P8.1) l
- Revision 24 of the Millstone Nuclear Power Station Emergency Plan corrected the {
previously identified concerns and provides an adequate planning basis for an acceptable state of onsite emergency preparedness in accordance with the ,
standards of 10 CFR 50.47(b), and the requirements of Appendix E to 10 CFR 5 (Section P8.3)
- The licensee conducted security and safeguards activities thoroughly and in a j manner that ensured safe operations, inspection were conducted in the areas of access authorization, alarm stations, communications, protected area access control of personnel and packages and protected area access control of vehicles. (Section S1)
- The licensee's security facilities and equipment in the areas of protected area assessment aids and personnel search equipment were determined to be well maintained and reliable. (Section S2)
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- The security force members adequately demonstrated that they have the requisite I knowledge necessary to effectively implement the duties and responsibilities j associated with their position. Security force personnel were being trained in j l accordance with the requirements of the Training and Qualification Plan and training i documentation was properly maintained and accurate. (Section S4)
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e The review of the licensee's audit program indicated that the security audits were comprehensive in scope and depth, that the audit findings were reported to the appropriate level of management, and that the program was being properly administered. In addition, a review of the documentation applicable to the self-assessment program indicated that the program was being effectively implemented to identify and resolve potential weaknesses. (Section S7)
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I Report Details Summarv of Unit 1 Plant Status Unit 1 remained in an extended maintenance mode for the duration of the inspection
- period. Personnel focused their efforts on performing corrective and preventative maintenance to maintain the plant in a safe shutdown condition, to minimize human performance errors,' and remain compliant with procedures, programs and policies, p1.1 Coerations !
U101 Conduct of Operations '
01.1 Blackness Testina of Soent Fuel Racks Inspection Scope (62707)
The inspectors observed the performance of blackness testing in the Unit 1 spent fuel poo Blackness testing was performed to measure thermal neutron attenuation in the wall of the Unit 1 spent fuel storage racks that utilize Boraflex as the neutron absorber material. The l
. objective of the measurement was to confirm the presence of the Boraflex absorber and ,
ensure it was intact. The testing was performed in storage racks 7 and 10. The testing was performed by a vendor and involved the use of a test tool containing a Californium -
252 source and four detectors. . As the test tool vertically traverses the selected storage cell, fast neutrons from the CF-252 source pass out through the walls of the cell, becom ' thermalized, and diffuse back towards the test tool. In the areas where the Boraflex is missing or significantly degraded, the neutrons pass through, and are absorbed by the~
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detector, and are registered as increased counts by the detector instrumentatio Increased thermal neutron count rates are interpreted as missing or degraded absorber material. The licensee committed to the NRC to perform blackness testing every two years to monitor existing or future Boraflex panel shrinkage and erosion to confirm that the criticality analysis assumptions for the spent fuel pool remain valid. The inspector also reviewed the " Blackness Testing" special procedure, SPROC 98-1-07, and the associated safety evaluation screenin <
1 Observations and Findinas The inspectors found that the work performed during blackness testing was completed in a professional manner. This was a significant effort by the plant staff that involved many organizations. Daily planning meetings and comprehensive pre-briefing were noted strengths for the evolution. Health Physics support for the activities was excellent including the control of foreign material exclusion areas. Work on the southeast corner of the fuel pool (rack #10) was completed without incident. However, the testing in the
- northwest corner of the pool (rack #7) identified issues that resulted in the work being
! prematurely terminate During performance of the testing in rack #7, the licensee identified travel stops installed on the northern extreme of the refuel bridge rail system, which restricted access to storage locations that were planned for inspection. This restriction made it necessary to change the SPROC 98-01-07,to include alternate destinations for double blade guide units which i
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were required to be moved to support testing. The .nspector discussed the configuration j control aspects of the rail stops with the licensee and a condition report (CR) was I generated. The inspector also discussed the appropriateness of removing specific guidance j for the blade guidc moves with respect to storage locations, during the SPROC revision !
l process. This issue was discussed with the Unit Director who indicated it clearly did not i meet management's expectations. Subsequently, specific guidance for the blade guide l l
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movement and storage locations were include in the SPROC revision prior to commencmg i wor '
After testing was resumed in rack #7, two of the four detectors in the test tool indicated a )
decline in sensitivity following the terting of the two initial cell locations. The vendor technicians suspected it was due to high background gamma radiation, and an underwater survey of storage rack #7 was performed. High gamma radiation was found in some locations, due to spent nuclear instrument tubes hanging on the wall of the spent fuel pool ,
adjacent to the cells being tested. This condition made it necessary to revise the SPROC i to include a camma survey for each of the remaining test locations prior to performing a !
test in that location. The licensee determined that this was a learning opportunity for ;
future blackness testing in that gamma surveys should be done prior to the contracted l testing, such that the effect of the gamma radiation on the datectors could be minimize Shortly after resumi g testing, a double blade guide unit became stuck on the main fuel hoist grapple. This made it necessary for the licensee to terminate the SPROC prior to the completion of desired data collection, having tested only two of the planned twelve cell locations in that rack. The licensee wi'l determine at a later date if sufficient data has been collected to satisfy the requirements of the SPROC. The inspector observed troubleshooting activities for the stuck blade guide at the end of the inspection period. The troubleshooting plan clearly defined the scope of the work with specific termination point During the pre-job briefing, responsibilities were assigned and contingencies were discussed. The troubleshooting was completed and a videotape recording was made of the activity for later review and for development of a plan to actually dislodge the blade guides frorn the main fuel hoist grappl The inspector attended a blackness testing critique following the completion of the wor All aspects of the worir were reviewed from the initial planning to the termination and removal of vendor equipment from the site. The majority of the plant staff that performed the work were present for the critique, as well as the Unit Director. Areas for improvement and examples of things that went well were discussed. A number of action items resulted from the critique including: a review of the present controls for material movement and storage in the spent fuel pool; a review of the restrictions on the refuel floor l general area lighting; development of specific operations guidance for placing Equipment in a safe condition following the identification of a problem; a review of the regulatory commitment to perform blackness testing; and a review of standards and focus for work on the refuel floor. The inspector noted an open exchange of information, with the Unit Director emphasizing that the standards are being raised for all activities on the refuel floo This was apparent to the inspectors during observations of this evolution. The licensee will review the commitment to the NRC and determine if that commitment was met in light of the early termination of the testin . _ _ _ _ _ _ _ .
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l Conclusions Blackness testing on the Unit 1 spent fuel pool storage racks to ensure the pr9sence and j integrity of the Boraflex panels was generally completed thorcughly and professionall i Conservative decision making and strong management oversight were apparent throughout the activities. All concerns raised by the inspectors were immediately addressed. Daily planning meetings and pre-job briefings were noted strengths for this evolution Health Phys;cs support, including foreign material exclusion control was excellen U108 Miscellaneous Operations issues (92700)
08.1 ACR M1-96-0564: Contractor Control issues (Updcte - SIL ltem 39) l Insoection Scope (9290_1)
The inspector reviewed the licensee's finding and corrective actions associated with adverse condition report (ACR) M1-96-0564, issued on September 13,1996, which decumented a negative trend with regard to the control of contractors. The Nuclear Safety Assessment Board (NSAB) identified the issue and requested that management perform a .
root cause evaluation and take appropriate corrective actions. The inspector rev%wed the licensee's closure package for this ACR, which included the root cause evaluation, action request closecut forms, copies of revised and new procedures for overseeing contractor activities, and the " improved Millstone internal Controls Action Plan." Observations and Findinas On November 18,1996, the licensee completed a root cause evaluation that provided an independent assessment of the breakdown of the overall contract process at Millston The assessment addressed various phases of the contract process inc!uding: contract planning; control and acceptance of work; demobilization of the workforce; and monitoring of the overall process. The assessment identified that the processes to control contractors at Millstone were very ineffective and validated the issue identified in ACR M1-96-056 The root causes for the ineffective program were poorly established management expectations, poor program guidance, and a weak process for advance planning and prioritization of wor The inspector reviewed the corrective actions associated with the findings of the root cause evaluation, in February 1997, the licensee issued the " Contractor Control Handbook," which provided guidance on roles and responsibilities for individuals who oversee contractor activities. Additional contract administrators were hired in 1997, and i the Purchasing and Contracts groups were merged to form a single entity, the Nuclear Contracts Administration. On January 28,1998, the Site Operations Heview Committee approved procedure OA 13, Revision 1, " Procurement and Administration of Contractor Services," which was written as a station level procedure to consolidate the requirements and processes involved from initiation of a contract for vendor services through the closure of the contract. In June of 1997, a self-assessment was completed, which concluded that corrective tions in the form of enhanced vendor control and management was being i
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implemented, but war, stilllacking the correct level of staffing to achieve management's i expectations, As a result of various interna' audits that identified deficiencies in administration of outside contracts, the Nuclear Controls Administration prepared the " Improved Millstone internal Controls Action Flan," in February,1998. The plan was developed to raise awareness of policies and procedures, define expectations for alllevels of personnel, and establish accountability when policies and procedures are violated. The plan also included the formation of a Steering Committee to monitor the success of the plan, Conclusions The inspector reviewed the corrective actions associated with ACR M1-96-0564 concerning a nagative trend with regard to the control of contractors at Millstone. These corrective actions included new procedures, a new organization for contract administration, l and on improvement plan with long term actions that are scheduled for completion at the (
end of 1998. SIL item 39, which addresses this ACR, will remain open pending '
completion of the improvement plan and a review of the effectiveness of the corrective ,
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.U.j.il Maintenance I
U1 M1 Conduct of Maintenance l M1.1 Diesel Generator Surveillance Insoection Scoce (6172Q) I l
i The inspector observed licensee performance during an emergency diesel generator (EDG)
surveillance which was accomplished in accordance with SP 668.1, " Diesel Generator Operational Readiness Demonstration," Observations and Findinqq
On May 28,1998, the licensee conducted a pre job briefing for an emergency diesel generator surveillance test which was attended by the applicable operating crew and i support staff. The licensee discussed the overall procedure, prerequisites and precautions, and personnel responsibilities, as well as various safety concerns and contingencies due to elevated smoke levels that have occurred in the EDG room in the previous two diesel L operations. In addition, other surveillance that were being performed either concurrently or after completion of the EDG operation were discussed during the crew brie The EDG was started and electrically loaded as directed by procedure, but approximately 30 minutes into the EDG run, the #3 cylinder indicator ( Kine) valve was opened to obtain peak cylinder firing pre ,sure measurements during performance of C.M.-102 " Diesel Engine Analysis," and subsequently seized while the indicator valve was being cycled in the closed direction. Conditioned based maintenance (C.M.) personnel could not successfully i
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close the valve, and as a result, the licensee made a conservative decision to secure the ED The licensee subsequently replaced the faulty #3 cylinder indicator valve, and successfully retested the EDG in accordance with SP 668.1 on May 29,1998. However, a few
minutes prior to the start of the EDG for <he second run, the inspector observed a '
technician cycle open and closed the #3 cylinder indicator valve that had been previously replaced. While C.M.-102 allows for operation of the indicator valves for obtaining the cylinder pressure measurement ar.d with the shift menager's permission, or through signature approval on the work order, the cyc{ing of the #3 cylinder indicator valve did not fall under thero categories. The inspector discussed the issue of personnel cycling indicator valves without specific procedurci guidance or operations approval, as well as working outside the scope of the work order with both the EDG system manager and the assistant operations manager. Consequently, a condition report was issued that detailed the C.M. technician's basis for working outside the scope of a work order and for cycling 1 the indicater valve without proper procedural or operations department directio The generic implications of the aforementioned indicator valve cycling, specifically configuration management, human performance and procedural complianco issues is considered unresolved (UBI 50-245/98-208-01)pending completion of the licensee's eva'uation of this even Conclusions ,
J The inspector concluded that while the licensea's perfortnance of the diesel generator surveillance was good, the generic implications of configuration management and human performance issues due to the #3 cylinder indicator valve operation without procedural or operations department direction is considered unresolved pending completion of the licensee's evaluation of this even U1 M2 Maintenance and Material Condition of Facilities and Equipment M2.1 (Uodatel UBI 50-245/97-02-02: Control and Calibration of Measurina and Test Eauioment Inspection Scope (62707)
The inspector reviewed the licensee's corrective actions for two condition reports (CR),
M3-98-0834 and M3-98-2864, concerning nonconforming measuring and test equipment j (M&TE) which may result in a condition adverse to quality. The concern focused on the ,
fact that if out-of-calibration M&TE was used to perform a surveillance on technical '
specification (T.S.) related systems, there was no requirement to perform a timely l evaluation of the impact with respect to operability of that equipmen l l \ Observations and Findinas in May of 1997, inspectors completed a review of the licensee's corrective action program and documented the results of that inspection in NRC inspection report 50-245/97-02, l r
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dated June 24,1997. During that inspection, an unresolved item (50-245/97-02-02)was opened concerning the interface between the corrective action program and other lower tier deficiency reporting processes. The concern was that these other reporting processes could circumvent procedure RP-4, " Corrective Action Program." The licensee initiated CR
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M197-1119 on May 14,1997, which identified multiple deficiency identification '
orograms, and it was not known whether they adequately interfaced with RP- One specific program reWewed by the licensee during CR M197-1119 follow up was the M&TE nonconformance report (M&TE NCR) system. M&TE NCRs at the time were governed by WC-8, " Control and Calibration o.f Measuring and Test Equipment," Revision 1. Under WC-8, a 30-day time frame was given from the time of discovery of an M&TE nonconformance to closeout of the issue. A CR would subsequently be generated during the evaluation process if a quality related measurement was determined to be affecte However, RP-4 states that conditions adverse to quality should be screened for Operability / Deportability within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery. Between September and December of 1997, the licensee reviewed this issue to determine the correct approach in handling M&TE NCRs.10 CFR 50 Appendix B, " Quality Assurance for Nuclear Power Plants," was reviewed by the licensee, along with the Quality Assurance Topical Report and the FSA In addition, M&TE control procedures were obtained from several nuclear stations, all of l which used some variation of the M&TE NCR system, with a required processing time )
between 5 and 30 days. The licensee revised WC-8 on December 10,1997,and i implemented a new 10-day requirement for M&TE NCR evaluation I On February 12,1998, the Unit 1 Experience Assessment Department initiated a CR (M3- l 98-0834), which documented that the allowance of a 10-day delay in resolving the I conditions reported by an M&TE NCR in accordance with WC-8 constituted a condition adverse to quality. Specifically, under certain circumstances, previous use of out-of- {
calibration M&TE could inadvertently affect the operability of safety-related plant si atems, structures, and components, and therefore, would require prompt review and report 1g to the control room within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery, per RP-4. The inspector reviewed tha licensee's follow up and closure of M3-98-0834. The CR was closed with no corrective action based on the fact that the current process was found to be an appropriate balance j between the need to make notifications of conditions adverse to quality and the desire to reduce the number of " false alarms" that needed to be reviewed by the shift manager. The CR stated that given the rigor of review for the creation of Revision 2 to WC-8, no change to the current process was warrante Subsequently, on May 5,1998, CR M3-98-2864,was initiated by the Director, Corrective Actions and documented a continuing concern with the inconsistency between RP-4 and WC-8 regarding the timeliness for evaluating nonconforming M&TE which may result in a condition adverse to quality. The inspector reviewed the corrective action plan for this CR i following the licensee's review and reconsideration of the issue. The licensee will revise i WC-8 to include the initiation of a CR immediately after a f aulty M&TE is identified as i having the potential to make the status of a structure, system, or component (SSC) !
indeterminate or inoperable. This will be accomplished by adding quality indicators to the j f M&TE " Custody Usage Log." For M&TE NCRs that do not affect quality SSCs, the 10-day evaluation period will apply. Additionally, Attachment 4 to RP-4 will be revised to read "A structure, system, or component status is indeterminate or inoperable as a result of f aulty
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l 7 M&TE screened in accordance with WC-8." These corrective actions will be completed by July 28,199 Conclusions The inspector reviewed the licensee's follow up and corrective actions for a concem associated with an inconsistency between RP-4 and WC-8 regarding the timeliness for
. evaluating M&TE NCRs which may result in a condition adverse to quality. The corrective actions were adequate to resolve this issue. However, UBI 245/97-02-02 remains open pending the licensee's review of the interface between RP-4 and other lower tier deficiency reporting processe U1 M8 Miscellaneous Maintenance issues M8.1 (Closedi LER 96-044-00 and 01: Hioh Ranoe Stack Noble Gas Monitor inoperable Due to inadeauate Calibration Inspection Scope (92700)
-The inspector reviewed the licensee's finding and corrective actions associated with the ~
surveillance calibration procedure for the high range stack monitor, which did not verify that the instrument signal output coNsponded to an acceptable response time limit. The issue was reported pursuant to teconicai specification (T.S.) 3.8.D.7, which required the
- submission of a Special Report to the NRC within 14 days if the high range stack noble gas monitor cannot be made operable within seven days. The inspector reviewed the licensee's closure package for this LER, which included engineering evaluations, automated work orders, associated ACRs and CRs, and revised procedures for radiation monitor calibrations. The inspector also verified that the licensee met the reporting requirements of T.S. 3.8. Observations and Findinas
. i On June 7,1996, the hcensee identified that the calibration procedure for the high range stack noble gas radiation monitor did not verify proper time response, as required in the T.S. definition of instrument calibration. The monitors were declared inoperable and alternative preplanned actions were initiated in accordance with 3.8.D.7. The cause of the
. evant was personnel error, in that the previous interpretation of T.S. calibration requirements failed to recognize the requirement to verify acceptable instrument response time. Since no specific response time was delineated for the high range stack noble gas ,
monitor in the T.S. or the UFSAR, the acceptance criteria and methodology for the I instrument calibration needed .to be developed. On July 19,1996, during the performance of the revised calibration procedure, the licensee discovered that the radiological calibration sources had been modified to the extent that proper calibration geometry could not be
- established. The original calibration sources were found to have been laminated in order to
! provide a more durable cov6 ring material. The licensee determined the add;tional famination thickness invalidated calibration geometry, and was reported to the NRC on August 29,1996, as a supplemental LER (96-044-01).
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i 8 The licensee has implemented corrective actions for these issues. . The calibration ,
procedure for the high range stack noble gas monitor has been modified to verify adequate l response time. The revised procedure, SP 406AA, " Stack High Range Radiation Monitor Calibration," Revision 10, uses a NationalInstitute of Standards and Technology traceable calibration source. The monitor was returned to an operable status on September 30, 1996. The historical and generic implications of the failure to verify adequate instrument response time during instrument calibrations were addressed separately by the licensee in i LER 96-043-00, dated July 3,199 I The events discussed in LER 96-043-00 were also reviewed and documented in NRC 4 l
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inspection report 50-245/96-06,Section U1.01.3 (unresolved item (UBI) 245/96-06-01). )
Both the UBI and LER 96-043 remain open pending completion of further NRC review and inspection.
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l Conclusions l The inspector verified that the calibration procedures for the high range stack monitor were !
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appropriately revised and concluded that the licensee's corrective actions were adequat l The closure package associated with these LERs was comprehensive and address all of the l licensee identified issues. LER 50-245/96-044-00and 96-044-01 are close l l
U1.Ill Enaineerina i
U1 E3 Engineering Procedures and Documentation l E Emeraency Diesel Generator Fuel Oil / Lube Oil Issues Inspection Scoce (92903)
l The inspector reviewed the commercial grade dedication (CGD) and receipt inspection activities as they relate to acceptance of fuel oil for emergency diesel generators (EDG) for all three Millstone units. In addition, the inspector reviewed the licensee's actions resulting from condition reports (CRs) concerning information Notice (IN) 96-67, " Vulnerability Of Emergency Diesel Generators To Fuel Oil / Lubricating Oil incompatibility." Observations and Findinos EDG Fuel Oil CGD and Receiot inspection
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The inspector ident'ified numerous examples of receipt inspection and commercial grade dedication activities concerning EDG fuel oil across all three Millstone units that were not documented in accordance with station procedures. Specifically:
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- Post-1995 maintenance receipt inspection reports were not being generated as required by station procedures, for the receipt of EDG fuel oi * In 1995, Nuclear Receiving Group initiated a change to the purchase order for the procurement of EDG fuel oil that removed their responsibility for performing receipt l
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inspections; the change order placed requirements for acceptance onto a common operating proceoure that identified fuel oil delivery sampling requirements and contained no reference to receipt inspection or commercial grade dedication requirements set forth by the previous purchase order. As a result, Quality Assurance receipt inspections were not being performed in accordance with station procedure *
In 1996, commercial grade dedication requirements were transferred to a computer-based system, Materials Information Management System (MIMS). The licensee could not produce documentation that would indicate commercial grade dedications were being performed and documented for EDG fuel oilin accordance with station procedures since the programs inceptio In NRC Inspection Report 50-245/98-207,a violation of Criterion Vil of Appendix B to 10 CFR 50, " Control of Purchased Material, Equipment, and Services," was documented. The violation stated that the licensee f ailed to assure documentary evidence that gas turbine generator (GTG) fuel oil conformed to the procurement requirements, and were avaliable prior to the use of the fuel oil. In addition, various procedural compliance issues were
. identified specific to the receipt inspection and commercial grade dedication activities o the gas turbine fuel oil. The licensee was also requested to address the generic implications raised by the violation as they relate to EDG fuel oil for all three Millstone units. The examples above are similar to those identified in the inspection report in support of the violation. (50-245/98-207-02)
EDG Fuel Oil / Lube Oilincomoatibility Information Notice (lN) 96-67, " Vulnerability Of Emergency Diesel Generators Fuel Oil / Lubricating Oil Incompatibility," was issued by the NRC on December 19,1996. This IN detailed EDGs at Calvert Cliffs nuclear station that had experienced excessive carbon deposition on the pistons and piston rings, as well as abnormal scuffing of cylinder liner A root cause analysis (RCA) team was assembled from various plant personnel, as well as personnel from the engine manufacturer (Wartsila SACM) and a consulting firm. The RCA team concluded the most probable cause of the failure mode for production of excessive deposits was the incompatibility of the tube oil with the lower sulfur content fue Subsequently, in September,1997, Nuclear Safety Engineering (NSE) initiated condition reports (CRs) for all three Millstone units based on the possibility the EDGs could become degraded by excessive fouling of the pistons and cylinders by ash deposits. The CRs were generated mainly as a result of evaluations performed by NSE due to IN 96-67, and l precipitated by the switch to low sulfur diesel fuel at the Millstone units without I corresponding adjustments to the diesel lube oil alkalinity (e.g., Total Base Number - TBN).
l The licensee evaluated the incompatibility issue for the Fairbanks Morse EDGs at Units 1 and 2, and the Colt Pielstick EDG at Unit 3 and concluded the following:
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- Based on the lube oil vendors' and EDG manufacturers' recommendations, no incompatibility between the fuel oil (with the lower sulfur content) and the tube oil I (with the same TBN number) exist *
The number of hours accumulated on the EDGs with the low sulfur fuel coupled with the performance data gathered from specific EDG parameters have not produced adverse indications consistent with the failures that occurred at Calvert Cliffs. Boroscope inspections and normal surveillance inspections have not identified any indication *
The incompatibility issue was specific to the design of the EDG at Calvert Cliffs, coupled with a high lube oil consumption rate and the use of a synthetic lube oil. In addition, no other failures have occurred in the industry aside from the failures at Calvert Cliffs, and mineral-based oils are used at Milleton *
Surveillance procedures (18 month interval) and engine diagnostic tools used during monthly diesel operation (i.e., Unit three EDG) will be used to accomplish cylinder and piston ring inspections to identify any indications similar to those identified at _
Calvert Cliff Conclusions While violations of NRC requirements have occurred relative to commercia! grade dedication and receipt inspection activities of emergency diesel generator fuel oil across all three Millstone units, no action is required pending the licensee's response to the similar violation contained within NRC Inspection Report 50-245/98-207.'In addition, the inspector concluded the evaluations performed by the licensee concerning the applicability of the lube oil / fuel oil incompatibility on the EDGs at all three Millstone units werts adequate to support their conclusion U1 E8 Miscellaneous Engineering issues E (Closed) LER 96-035-00/01/02: Leakina Gamma-Radicaraohv Access Port Plua in
- the Low Pressure Coolant Iniection (LPCI) Syste!p Pipina Inspection Scope (92700)
The inspector reviewed a closure package for an LER submitted by the licensee due to the identification of a leaking gamma radiography access plug in the LPCI system. The revlew also consisted of interviews, evaluation of the licensee's response in accordance with 10 CFR 50.73, as well the review of work orders and other documentation contained within the closure package in support of the closure of the LE Observations and Findinas On May 6,1996, the licensee identified a leaking gamma radiography (gamma port) access plug in the LPCI system piping. The licensee determined from ultrasonic testing (UT) that a t_ _ _ _ _ _ _
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l crack, characteristic of intergranular stress corrosion cracking (IGSCC), was the source of l the leak. Subsequently, inspections of other gamma ports identified one additional l cracked plug also in LPCI system piping. The licensee subsequently issued LER 96-035 l (followed by revision one and two) as a reportable event pursuant to 10 CFR 50.73. The licensee established four commitments as corrective actions for the event: (1) to repair the 1 two identified cracked gamma plugs by weld overlay or pipe replacement prior to startup l for cycle 16; (2) to incorporate gamma plugs into future IGSCC in-service inspection (ISI)
program prior to startup for cycle 16; (3) to issue a supplement to LER 96-035 when metallurgical results have determined the cause of the gamma plug cracks; and (4) to review other small bore piping (less than four inu i m diameter) welds for possible l
inclusion in the IGSCC ISI program if found to be susceptible to IGSC The inspector verified completion of the commitments contained within the LERs, in that:
(1) LPCI piping was replaced that contained one of the cracked gamma plugs (and also contained a gamma plug that had passed UT inspection), and the second gamma plug was repaired by the weld overlay method; (2) the 15 remaining gamma plugs were incorporated into the licensee's lGSCC ISI program; (3) the cracked gamma plug contained in the section of piping removed from the LPCI system was tested and the crack was verified to have been caused by IGSCC; and (4) the licensee reviewed small bore piping welds for susceptibility to IGSCC and for possible inclusion in the IGSCC lSI program. The licensee '
concluded that the inclusion of the remaining small bore piping welds in the IGSCC program was not warrante The inspector verified that the licensee's response to the event was in accordance with the provisions of 10 CFR 50.7 C_o. nclusions The inspector verified that the licensee's corrective actions have been completed and are adequate to support closure of the LER. LER 50-245/96-035-OO/01/02is close l l
E8.2 (Closed) LER 96-013: Unanalyzed Condition due to Indeterminate Boraflex Degradation Inspection Scoce (92700_) )
l The inspector reviewed a closure package for an LER submitted by the licensee due to the
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identification of gaps in Boraflex panels, a neutron absorber used in storage racks in the spent fuel pool. The inspector conducted interviews, evaluated the licensee response in accordance with 10 CFR 50.73, and reviewed other documentation contained within the closure package in support of the closure of the Ld i Observations and Findinas j l
On February 20,1996, the licensee determined an unanalyzed condition existed due to the l presence of gaps that were identified during " blackness testing" performed in September, i 1995. Blackness testing was performed to verify the continued presence and effectiven3ss of the Boraflex panels utilized in the spent fuel pool storage racks, such that
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I 12 l the existing criticality analysis which bounds the storage of spent fuel could be validate However, while the gaps were determined to have a negligible impact on the criticality
. analysis, the licensee questioned whether the projected growth rate of the gaps could be 1 adequately predicted to ensure the existing criticality analysis would remain valid. The I licensee reported this condition as an unanalyzed condition in accordance with 10 CFR 50.72. The licensee subsequently performed a safety analysis utilizing the criteria ,
contained in 10 CFR 50.59, and determined an unreviewed safety question existed due to a malfunction (Boraflex gap formation) of a different type than previously evaluated. While ]
l the licensee determined that no safety consequences existed due to this event, LER 96-013 was submitted to document and report the unanalyzed condition.
I The inspector verified that corrective actions documented in the LER were completed to support continued operability of the spent fuel pool racks, and to maintain compliance with the technical specifications (T.S.). Specifically, the licensee: (1) established procedural provisions to ensure a one year delay period for irradiated fuel was instituted prior to the >
placement of fuelinto racks containing Boraflex panels, due to the established effects of L
gamma radiation on the initiation and advancement of shrinkage and gaps in Boraflex; (2)
l performed additional blackness testing to verify the effectiveness of the Boraflex fuel racks l .to maintain a subcriticality margin of 10% (Keff of .90 or less), thereby ensuring compliance with T.S. limits; and (3) submitted a T.S. change request to tlie NRC to change the Km (K-infinity) fuel design limit in the normal reactor configuration at cold conditions to l a maximum of 1.24.
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, The NRC subsequently issued License Arnendment 97, which supported the licensee's -
! submittal of a more restrictive fuel reactivity limit (maximum Km of 1.24)in the spent fuel I pool to ensure the Keff T.S. limit of 0.90 is not exceeded. The licensee's T.S. change submittal documented a margin of safety ensured not only by analyses that were bounded l by conservative Boraflex gap sizes, but also by a commitment to continue blackness
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testing to verify the continued presence and effectiveness of the Boraflex panels in the spent fuel.
l The inspector verified that the licensee's response to the event was in accordance with the L provisions of 10 CFR 50.73, Co_ nclusions l The inspector verified that the licensee's corrective actions have been completed and are j adequate to support closure of the LER. LER 50 245/96-013is closed.
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Report Details Summary of Unit 2 Status Unit 2 entered the inspection period with the core off-loaded. The Facility 1 service water system header has been drained for the entire report period to support system repair activities. The unit was initially shut down on February 20,1996, to address containment sump screen concerns and has remained shut down to address the problems outlined in the Restart Assessment Plan, an Order requiring an Independent Corrective Action Verification Program, an Order imposing a requirement to develop and implement an effective employees concern program and an NRC Demand for Information [10 CFR 50.54(f)] letter requiring an assertion by the licensee that future operations are conducted in accordance with the regulations, the license, and the Final Safety Analysis Repor U2.1 Operations U201 Conduct of Operations 01.1 General Comments (71707)
Using inspection Procedure 71707,the inspectors conducted frequent reviews of ongoing-plant operations. The inspectors noted good sensitivity to special evolutions and equipment outages that affected shutdown safety. Specifically, short-duration outages of
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both Unit 2 emergency diesel generators to support continuing service water system work *
were well controlle U2 08 Miscellaneous Operations issues (92700)
08.1 IGjosed) Escalated Enforcement item 50-336/96-06-05: Reactivity Controls Durina Plant Cooldown (Closed - Unit 2 Significant items List No. 9 Individual item)
(92702) j a. Insoection Scoce The inspector reviewed licensee corrective actions associated with Escalated Enforcement item (EEI) 50-336/96-06-05,which involved two final safety analysis report (FSAR)
discrepancies.
! Observations and Findings i
The two discrepancies concerned the initiation of cooldown prior to attaining the boron concentration necessary for cold shutdown and the performance of the cooldown evolution without maintaining the shutdown group of control rods fully withdrawn. This issue was reviewed in NRC Inspection Report 50-336/98-206. The report stated that the safety evaluations to support the needed FSAR changes were reviewed by the NRC to support Technical Specification Amendments 116 and 133, but the issue remained open because the FSAR had still not been updated to reflect the amendments. During the current l
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inspection period, the inspector reviewed the FSAR change, which was approved by the Unit Director on May 11,1998, and found the change adequately addressed the discrepancie Conclusions eel 50-336/96-06-05is considered close U2.ll Maintenance U2 M1 Conduct of Maintenance M1.1 General Maintenance Observations Inspection Scope (61726)
Through inspection, ascertain that surveillance of safety significant systems and components are being conducted in accordance with technical specifications and other requirement _ Observations and Findinos During routine plant inspection tours, the inspectors determined that the maintenance and surveillance activities observed on a random sample basis were being properly performe Interviews of licensee field personnel were conducted to verify the adequacy of work controls. The inspector observed all or part of surveillance activities performed under the following procedures:
- SP 2601C Boric Acid and Chemical and Volume Control System Valve Tests
- SPROC 97-2-14 HPSI Injection Loop Flow Instrumentation Verification Test
- SPROC 98-2-03 HPSI discharge Valve Dynamic Testing
- SPROC EN98-2-10 LPSI System Pump Start Dynamic Test Conclusions The inspectors found the testing performed under these procedures was characterized by good communications and strong control of individual evolutions. The operators and test directors demonstrated conservative decision-making by halting testing when unexpected
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conditions were encountered. The conditions were promptly dispositioned and corrected when appropriate. The testing demonstrated acceptable performance of the subject components
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15 l U2.lli Enaineerina U2 E1 Conduct of Engineering E Control and Use of Vendor Information (92701 )(Updated - Unit 2 Significant items List No. 50) - Inspection Scooe The inspector reviewed the status of the licensee's Vendor interface program for Unit 2 in i accordance with NRC Generic Letter (GL) 90-03, " Relaxation of Staff Position in Generic i Letter 83-28, Item 2.2 Part 2 ' Vendor Interface for Safety-Related Components'."
~ ' Observations and Findinas The overall Millstone site program for vendor interface was initially reviewed in NRC
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Inspection Report (IR)97-203. The review was performed at both the site and individual unit levels in order to verify that appropriate policies and procedures were in place and i
, _were being effectively implemented. Many aspects of the licensee's programs to addres the GL 90-03 items were site level programs that are being applied to all units. All of the i common site aspects and the Unit 3 specific aspects were reviewed, but the unit specific I aspects for Units 1 and 2 were not all fully reviewed at that tim The common site aspects were either documented as acceptable in NRC IR 97-203 or, when areas were noted to have problems, were subsequently upgraded by the licensee and documented as satisfactory in NRC IR 98-207. For Unit 2, the remaining area for review of GL 90-03 items is the unit specific implementation of procedure DC 16, " Vendor Equipment Technical Information Program (VETIP)," in the following areas: Key Safety Related Equipment List (KSREL), vendor contacts, vendor manual updates, and procedure update .The inspector reviewed Technical Evaluation M2-EV-98-0038, Rev. O, " Key Safety Related Equipment List, Components Selection at Millstone Unit 2," dated February 5,1998. This was reviewed against procedure DC 16, GL 90-03, and the Millstone Unit 2 System List, Revision dated May 28,1998. There are 11 equipment categories,23 plant systems, and l 450 components contained on the KSREL. These are addressed by 30 vendor manual The inspector noted a few additional systems that appeared as if they should also be
' included on the KSREL. The licensee provided acceptable justification for their exclusio The licensee provided the following status of the vendor contact activities and the vendor manual update process: twelve manuals have been fully updated and approved, five have been updated and are in the approval cycle, and thirteen are in the update process. The inspector selected a sample of the manuals and reviewed them against the criteria of procedure DC 16. Discussions were held with the personnel responsible for the manual update process. Some of the older vendor manuals are being consolidated as the revision ;
process proceeds. The licensee stated that the KSREL would be updated as the manuals l are completed in order to cross reference the appropriate revised manuals and add support !
components that are related to other components already in the KSRE i
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16 i The procedure reviews against the updated manuals have not yet been completed. All work is scheduled for completion before startu I Conclusions i The licensee's Vendor Interface Program for Unit 2 is in accordance with NRC Generic Letter (GL) 90-03 for all areas reviewed. The remaining areas (vendor contacts, vendor manual updates, and procedure updates) were progressing along an acceptable path, but i were not complete. Therefore, Significant items List No. 50 remains open pending completion of these areas and is considered update E1.2 Final Safety Analvsis Report (FSAR) Updates (37001)(Updated - Unit 2 Significant l ltems List No. 2) ! Scoce of insoection A selected sample of ten plant design and FSAR changes were chosen for a detailed review to evaluate the effectiveness of procedure NGP 3.12, " Safety Evaluations," and
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procedure RAC 03, " Changes and Revisions to Final Safety Analysis Reports." _
j Observations and Findinas The plant design and FSAR changes were reviewed to evaluate the licensee's conformance with procedural requirements, the adequacy of the screening process in addressing the unreviewed safety question criteria specified in 10 CFR 50.59, the adequacy of technical information conteined in the supporting safety evaluations, and the consistency between i the level of detailin the FSAR changes and the overalllevel of detail in the updated FSA l The inspector determined that the changes reviewed were consistent with the procedural guidance,10 CFR 50.59 was adequately addressed, the safety evaluations edequately supported the changes, and the FSAR updates sufficiently described the changes. No unreviewed safety questions were identified by the inspecto Conclusions
The inspector concluded that the licensee's procedure NGP 3.12 and procedure RAC 03 have been effective in that the design changes and the FSAR changes were adequately !
documented. Because this inspection is one of severalinspections assessing the design
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and FSAR change processes and their effectiveness, the NRC's final determination of the overall adequacy of the licensee's 10 CFR 50.59 safety evaluation and FSAR change processes will be documented in a future inspection report. Therefore, Significant items List No. 2 will rernain open and is hereby update E1.3 Procedurally implemented Temocrary Modifications 1 Insoection Scope (37551)
The inspectors reviewed the implementation of temporary modifications specified in approved plant procedures. The inspection included review of station administrative l
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procedures and operating procedures that specified installation of temporary modification The inspector also interviewed engineering and operations personnel, and cognizant procedure owners, ObsqryJtions and Findinos i
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The licensee defmed temporary modification as a change to the original design configuration of a structure, system, or component that will be restored to the original design configuration or modified permanently. Procedure WC10, " Temporary Modifications," provides the instructions for review and control of temporary rnodifications. 1 However, procedure WC10 exempted temporary modifications that were installed and {
controlled by approved plant procedures from the review and control provisions of I procedure WC10. A condition report (CR), CR M3 97 4556, documented concems related I to the adequacy of design reviews for temporary modifications implemented through approved procedures rather than through the process described in procedure WC1 I To evaluate the current adequacy of reviews for temporary modifications implemented through approved plant procedures, the inspector reviewed procedures DC1,
" Administration of Procadures and Forms," DC2, " Developing and Revising Millstone I Procedures and Forms," and DC3, " Verification, Validation, and Approval of Procedures and Forms, " which, collectively, provico the administrative controls for creating procedural instructions. Although curnbarsome and subject to personnel errors at several decision points where the scope of review is established, the inspector concluded that these administrative procedures directed the performance of appropriate safety evaluations and I design reviews for temporary modifications implemented through approved procedure Nevertheless, the licensee established a corrective action plan for CR M3-97-4556to perform the following: (1) revise procedures DC2, and DC3 to ensure that future temporary modifications incorporated in procedures receive reviews comparable to reviews for temporary modifications implemented directly through procedure WC10, (2) revise procedure WC10 to provide a process for review of temporary modifications incorporated in procedures, and (3) revise procedure DC3 to require that temporary modifications already incorporated in procedures receive an engineering revie The licensee hplemented Change 5 to procedure WC lO, Revision 1, on April 15,199 The inspector found that this change to procedure WC10 introduced a process to review temporary modifications incorporated in procedures that was similar to the review process for temporary modifications implemented directly through procedure WC10. The inspector determined through interviews with the cognizant procedure owner that the licensee plans to change procedurcs DC1, DC2. and DC3 to refer to the process in p ocedure WC10 for review of temporary modifications incorporated in approved procedure The inspector also examined the method of ensuring adequate design reviews of existing ( temporary modifications implemented through approved procedures. The mechanism for l periodic review of procedures was the licensee's biennial review process defined in procedure DC3. The cogrizant procedure owner stated that the planned corrective action to conduct an engineering review of existing temporary modifications incorporated in procedures involved changing the biennial review process in procedure DC3 to include the following actions: (1) identify temporary modifications incorporated in procedures during i
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the biennial review and (2) conduct the review specified in procedure WC'l0 for temporary modifications implemented by procedur The licensee's planned corrective action was tc, review the existing temporary modifications incorporated in procedures during each procedure's biennial review. This does not ensure review prior to the next implementation of the temporary modification, nor does it provide for review of currently installed temporary modifical;ons incorporated in procedures. The inspector discussed this issue with the Unit 2 Engineenng Director and !
the Operations Manager. As an interim corrective action, the licensee issued an operations )
night order precluding implemeritation of temporary modifications incorporated in procedures without an engineering review. The operations department documented this )
concern in CR M2-98-1866 for development of long-term corrective action l The inspector reviewed the following three temporary modifications that the licensee had implernented through operating procedures: (1) installation of alternate cooling to the "B" wasts gas compressor using section 4.16 of procedure OP2330A. "RBCCW System,"
which was installed under Automated Work Order (AWO) M2-98-01991,(2) installation of l
alternate power to control roem lighting panel L77 using section 5.19 of procedure OP2344A, "480 Volt Load Centers," which was installed and removed under AWO M2-97-08037, and (3) installation of jumpers to block the safety injection actuation signal (SIAS) {
to each of the four low pressure safety injection (LPSI) system discharge valves using section 4.7 of procedure OP2207," Plant Cooldown," which were installed under AWOs M2 96-01437/38/39/40. These recurring temporary modifications were not described in I the Millstone Unit 2 Final Safety Analysis Repor l The inspector determined through interviews that the temporary modifications implemented through these procedure sections had not received design reviews to verify the adequacy of the design changes. Although the procedure sections had been subject to safety evaluation screening, that type of screening would not address the following potential concerns with implementation of the temporary modifications: the pressure rating of hoses ,
relative to service conditions, the compatibility of materials, the coordination of electrical protective devices to isolate as close as possible to faults, and the response of control circuits to input signals. The safety significance of these concerns was low because the licensee has perforrned an engineering a.ssessment of the temporary modification to the RBCCW system, the licensee had removed the alternate power supply for control room lighting, and the shutdown cooling system has not been in servic Conclusions The licensee identified concerns with the review process for temporary modifications implemented through approved plant procedures and implemented a corrective action plan to address future procedure changes that include temporary modifications. However, at
- Millstone Unit 2, the licensee f ailed to fully address concerns with existing temporary ;
l modifications implemented through approved plant procedures. This item is unresolved pending the licensee's completion of corrective actions to ensure that temporary modifications previously implemented through procedures are timely reviewed ( UBI 50-336/98-208-02)
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! U2 E8 Miscellaneous Engineering issue:
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t ( E8.1. . (Closed) LER 50-336/96-33: Incorrect F_ynctional Settinas on Steam Generator l Hydraulic Snubbers (9270_0_) Insoection Scope (9270_0)
l' The scope of this inspection was a review of corrr,ctive actions taken to address the l concerns discussed in Licensee Event Report (LER) 50-336/96-33.
! Observations and Findinas
l LER 50-336/96-33 discussed issues conceming the current lockup and bleed rate settings j for the steam generator snubbers that do not conform to their original design basis. The )
l bleed rate values will not support the design function of these snubbers to limit i displacement of the steam generators under a main steam line break (MSLB), which could -
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result in exceeding design allowable stress levels for the reactor coolant system (RCS)
piping. The affected snubbers are 10 inch bore, Figure 200,5 inch stroke, hydraulic snubbers manufactured by ITT Grinnell. The licensee determined that the cause of this l event was a breakdown in the design control processes used for the steam generator -
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hyc8raulic snubber The first corrective action specified in the LER was to reevaluate the design basis for the -
steam generator hydraulic snubbers, implement plant modifications to ensure that the snubbers will perform their intended function, and generate appropriate design documentation which willinclude values for steam generator snubber settings. The inspector reviewed Vendor Calculation N-ME-C-021, Rev. 02,10/19/97," Millstone Unit 2 -
Relief to Steam Generator Snubber Functional Requirements." There are eight direct acting hydraulic snubbers for each of the two steam generators. The snubbers are oriented ,
parallel to the hot leg axis, and are required to allow thermal movement of the steam J generators during normal operation, but resist dynamic movement from a seismic event or an MSLB. When subjected to the rated load of design basis earthquake (DBE), originally specified as 225 kips, the snubber rods were originally specified to take at least 30 seconds to reach a maximum displacement of 0.375 inches. The minimum time to reach a maximum displacement is referred to as the maximum bleed rat To accommodate thermal growth of the piping systems during normal plant operation, a minimum lock-up velocity, originally specified as 1.0 inch / minute, was established. Since the snubbers must activate to support the steam generators and restrict their movement ,
under dynamic conditions, a maximum lock-up velocity, originally specified as 2 l inches / minute, was establishe i
In the referenced calculation, the maximum b.eed rate was determined to be at least 30 seconds to reach a revised maximum displacement of 0.500 inches, as opposed to the original 0.375 inches. The minimum lock-up velocity was unchanged. The maximum lock-up velocity was increased to 1212 iraches/ minute, with a further restriction that all of the snubbers on any single steam generator must have maximum inck-up velocities within 4 inches / minute of each other. The required snubber faulted load capacity, the one-time
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combination of DBE and MSLB loads, was decreased from 437.5 kips to 300 kips. The preparer, ABB, used the Combustion Engineering-developed SGN-lli computer code to l calculate the dynamic fluid loads as a result of a main steam line break and a main feedwater line break. This code was approved for use by the NRC in Section 6.2.1.4 of the "NRC Standard Review Plan," NUREG-0800, Rev.1, July 198 t l To implement the revised lockup velocities and bleed rate for the steam generator hydraulic !
snubbers, the licensee issued Design Change Request (DCR) M2-97020," Steam Generator Snubber Control Va;ve Upg ade," Rev. 00,06/13/97. This DCR covers the functional l design basis for the steam penerator snubbers and the hardware modifications to restore the design basis. The snubers are required to limit the displacement of the steam -
generators under a seism and/or postulated main steam line pipe rupture accident to .
maintain the design stress limits for the attached RCS piping. The control valves of these hydraulic snubbers govern their functional operation and are therefore, the critical '
component relative to design basis compliance. Prior to the changes implemented by the DCR, the control valves that were in place had been installed in 1985 and due to their design configuration, did not provide adequate adjustment or control to maintain the required bleed rate setting. Under the DCR, new control valves were purchased and I installed to repbce the previous control valves in order to restore design basis compliance for the hydraulic snubber At the time of this inspection, the remaining open items identified by the licensee to complete the first corrective action listed in the LER were as.follows: (1) The control valve replacement has been completed on 14 of the 16 steam generator snubbers. Access to the remaining two snubbers is currently restricted by the placement of the missile shiel The inspector verified that the licensee is tracking the replacement of the remaining two f snubbers (Action Request 96034250-04);and (2) Update and close-out of DCR M2-97020, as well as update of the Main Steam System Design Basis Summary will be done l J
when all field modifications are complete and the vendor provided maintenance procedure for these snubbers has been received and entered into the plant documentation file Based on the licensee's progress thus far and their tracking of the remaining work, the inspector found the licensee's actions to address the first corrective action listed in the LER to be acceptabl The second corrective action specified in the LER was to review all other plant hydraulic snubbers to ensure that the design bases are adequately reflected in design and test documents and complete any necessary actions required as a result of the review. The licensee initiated five calculations to cover the 131 safety-related hydraulic snubbers, i excluding the 16 steam generator hydraulic snubbers.. One calculation provides relaxed I test acceptance criteria for the 131 safety- related hydraulic snubbers by increasing the ranges for the snubber lock up velocities and bleed rates when testing at 70 F ambient temperatur The results of the calculations were translated into Engineering Procedure SP 21174, j l " Snubber Functional Testing," Rev. 7, effective 07/25/97. This procedure provides the '
l sample selection basis, test frequency requirements, test criteria and the documentation requirements for all hydraulic snubbers, including the steam generator snubbers. The allowable test ranges for the lock-up velocities and bleed rates, as applicable to the steam l
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i generator snubbers, are provided for all hydraulic snubbers individually. Prior to the !
issuance of this procedure revision, the snubber test requirements were identified
l generically for groups of snubbers. The licensee considered the implementation of the '
revised procedure a major enhancement to the snubber testing program which will minirnize the possibility of loss of design basis control for hydraulic snubbers in the futur The inspector considered the licensee's corrective action to be acceptable.
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i The third corrective action specified in te LER was to review the design control processes l l
associated with this event to identify rieficiencies that could have contributed to this event I and complete any necessary actions required as a result of the review. The licensee identified that initial functional requirements for the 16 steam generator hydraulic snubbers f
l were established by the Combustion Engineering purchase specification and verified by l testing prior to shipping in 1971 by the manufacturer Grinnell. In Combustion Engineering Specification No. 18767-487-503," Project Engineering Specification for a Hydraulic Snubber System," Rev.1, issued 09/08/71, Section 4.0, the steam generator snubber {
bleed rate was specified to be at least 30 seconds to reach a maximum displacement of '
O.375 inches. In Section 6.0 of this specification, the lock-up velocities were specified as a minimum of 1 inch / minute and a maximum of 2 inches / minute in both compression and tensio During the mid-1970's, significant problems developed with these snubbers and studies were conducted to document the acceptability of greater tolerances on the operability range for hydraulic snubber control valves. An industry wide effort was undertaken to l
address and correct faulty Grinnell control valves and to increase the acceptable operating !
range for the Grinnell control valve Since the steam generator hydraulic snubbers had been improperly set since 1977 to a bleed rate and lock-up velocities outside of the original design requirements established in i the 1971 Combustion Engineering project specification, LER 50-336/96-33 was formally I submitted to the NRC on November 25,1996. Settings for all other hydraulic snubbers I were not outside of their design ranges and so were not reportabl l
The licensee issued memorandum MP2-DE-97-126," Closure of A/R 96034250-3, ;
' Evaluation of MP2 Design Control Processes.'" dated February 25,1997. The memo l refers to Adverse Condition Report 8671," investigation of Millstone 2 Configuration Control Deficiencies," as identifying that the root cause of the design basis discrepancies originated prior to 1990 and resulted from tow standards and inadequate design control l processes. Currently, the comprehensive Design Control Process under the site-wide :
Design Control Manual ensures that adequate engineering reviews are conducted prior to making plant design modifications. Also, the Configuration Management Project reviews were intended to capture alllicensing and design basis criteria for the plant to ensure consistency between plant configuration and the supporting design documentation. The licensee concluded that the noted programs, together with higher standards and an increased awareness of the significance of design configuration control will adequately j minimize any future occurrence of this concern. The inspector considered the licensee's actions satisfactorily addressed the third correctise action specified in the LER.
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22 G.onclusions The NRC concluded that the licensee's corrective sctions were acceptable to address the concerns described in LER 50-336/96-33 which involved incorrect functional settings on the steam generator hydraulic snubbers. This condition could have resulted in excsading design allowable stresses of the reactor coolant system piping during postulated event LER 50-336/96-30is close E .l. Closed) LER 50-336/97-17:Bypasseq Refuelino Machine Overload Cut off In_soection Scope (92JOO)
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The scope of this inspection included a review of Licensee Event Report (LER) 50-336/97-1 Obf3rvations.pndfindinay, LER 50-336/97-17 discussed that the technical specification (TS) requirements for the overload cut offs for the refueling machine hoist was not met in that the overload protection was automatically bypassed for approximately nine inches of vertical travel as it transitioned from the " fuel only" region to the " fuel plus hoist box" region. An overload cutout is provided on the hoist while the fuel assembly is being raised to protect the fuel assembly from excessive lifting force in the event the fuel assembly becornes inadvertently engaged with the core internals and pressure vessel during the lifting operatio Within the refueling bridge mast assembly is the hoist box which is moved vertically using the hoist cable. To remove a fuel assembly from the core, the refueling bridge mast is aligned over the desired fuel assembly and the hoist box is lowered using the hoist cabl At this point the hoist box assembly is suspended on downstops, which are attached to the mast, and the weight of the hoist box is no longer being carried by the hoist cabl The hoist cable then lowers the grappling device which is engaged to the top of the fuel assembly. Then the fuel assembly is raised out of the core region and into the hoist box assembly, During the first part of its tevel, the hoist assembly lifts only the weight of the grapple device and fuel assembly. Technical Specification 3.9.6.b requires an overload cut off of less than or equal to 1700 pounds in this " fuel only" region. When the fuel ass.embly is fully withdrawn within the hoist box, the hoist cable will begin to pick up the I weight of the hoist box in addition to the grapple and fuel assembly. The fuel assembly l
and hoist box is then raised into refueling machine mast so that the fuel assembly can be j
- transported. Technical Specification 3.9.6.d requires an overload cut off of less than or t l equal to 3000 pounds in this " fuel plus hoist box" region. However, the licensee found that the overload load cut off was bypassed in the area of the hoist box downstop as part
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of the original plant design to prevent the additional loads from the inertial and frictional l forces from exceeding the overload setpoint and shutting off the hoist motor. The licensee learned of the technical specification discrepancy after it was discovered at another facilit LER 50-336/9717 specifies that the root cause of this event was the failure to incorporate refueling machine design considerations (overload bypass function) during development of
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the original technical specification criteria. The corrective action specified in the LER was to modify the refueling machine hoist to eliminate bypassing the overload cut off or to
. change the technical specifications to reflect the bypass.. The licensee decided to modify the refueling machine by eliminating the bypass and shifting the overload cut off setpoint from the " fuel only" setpoint to the " fuel plus hoist box" setpoint at a mast height of 126.16 inches, one inch below the downstop of 125.16 inches. Shifting the setpoint one inch below the downstop is necessa.y because the hoist begins to encounter frictional and inertial forces in this one inch of travel which would cause the hoist to reach the " fuel only" setpoin The inspector evaluated whether shifting the setpoint one inch below the downstop rather than at the elevation of the downstop satisfied the technical specification requirement Neither the final safety analysis report nor the technical specifications define the precise elevation of the " fuel only" and " fuel plus hoist box" regions and they do not specifically state that the transition occurs at the height of the downstop. Therefore, the licensee's determination that the transition occurs somewhere between 126.16 inches and 125.16 inches when the hobt begins to encounter frictional and inertial forces was found to be acceptable. Therefore, the licensee's decision to shift the setpoint one inch below the downstop was considered acceptabl l The modification to eliminate bypassing the overload cut off and establishing the setpoint shift at one inch below the downstop has been completed and is awaiting retest. The retest, as well as operator training associated with the modification, is scheduled to be --
perforrned during the preoperational checks of the refueling bridge prior to core reloa Conclusion The inspector concluded that the licensee's corrective actions for this LER were acceptable and therefore, LER 50-336/97-17is closed. Licensee performance was good in that this condition was identified by reviewing events at other facilities. The bypassing of the i overload cutoffs on the refueling bridge for nine inches of travelis a violation of TS 3. ,
This non-repetitive, licensee-identified and corrected violation is being treated as a Non- i Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Polie (NCV 336/98-208-03) l E8.3 LClosed) LER 50-336/97-30: Weld Failure on "A" Emeraency Diesel Generator Lube i Oil Pioina
, Insoection Scoce (92700) l The scope of this inspection was a review of corrective actions taken to address the l
concerns discussed in Licensee Event Report (LER) 50-336/97-30.
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1 24 Observations and Findinas l
Condition Report (CR) M2-97-1536 documented a lubricating oil (LO) leak on the "A" emergency diesel generator (EDG) caused by a failed weld joint in the 4-1/2 inch diameter discharge tubing of the motor driven LO pump. The failure occurred after approximately one hour of EDG operation, following an unplanned start on August 2,1997, and resulted in a five to seven gallon LO spill. The "A" EDG was subsequently declared inoperable, and
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the portion of failed tubing was removed for evaluation and repai The preliminary examination of the 3 inch circumferential crack (Non-Conformance Report 297-279) concluded that the failure was due to fatigue of the weld itself which "had a significant amount of lack of weld penetration" varying between 30% to 50% of the pipe wall thickness. A further examination of the removed tubing segment revealed that this
" lack of penetration" existed at other welds and was likely present throughout the remainder of the LO piping system. The licensee referred to the original Purchase Order Specification 7604-M-160 and correspondence from the manufacturer to identify which welding codes or practices were applied during fabrication. In a letter from the manufacturer to the licensee's agent, dated February 8,1971, the manufacturer stated that " ANSI B-31.1 for piping ....is not applicable to the skid piping" and that " Skid piping I conforms to [ manufacturer's] standards and is built with space constraints as well as satisfactory service being the principal criteria." l When subsequent inspections revealed that partial penetration welds also existed in the
"B" EDG skid large bore piping, both the "A" and "B" EDGs were declared inoperable (Operability Determination (OD) MP2-044-97, dated August 27,1997). The licensee determined that this concern was reportable as a condition that was outside the design basis of the plant. The corrective actions described in LER 50-336/97-30 included reworking partial penetration welds in large bore piping and tubing to restore them to code acceptable full penetration welds prior to entering Mode 4. Following the weld repairs in the most critical Isrge bore lines, both EDGs were declared operable, but not yet fully qualified (Final OD MP2-044-97, dated September 6,1997). Additionally, skid-mounted small bore piping for both EDGs was reviewed and found operable with no modification <
(OD MP2-055-97, dated September 30,1997).
With the issuance of the LER, the licensee also issued Condition Report (CR) M2-97-1850 to document and correct the dieselissue. As corrective actions: (1) The welds on the large bore piping on both diesels were reworked and restored to full penetration welds; (2)
The "A" and "B" EDG large bore, skid piping was analyzed for both the "as found" and the
" reworked" configurations; and (3) The licensee established and implemented a vibration monitoring program, with acceptance criteria, to monitor the skid mounted piping. With completion of all corrective actions, the "A" and "B" EDGs were found operable and fully qualifie The inspector reviewed the licensee's analyses that evaluated the seismic integrity of the
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large bore piping as it existed prior to any weld repairs. The analysis of the partial penetration welde showed that although the predicted dynamic stresses were below the allowable stress levels, when the static stress induced by internal pressure was added to the dynamic stress component, a gross overstress condition resulted. The stresses ,
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induced by pressure were seen to be particularly large because of the presence of slip couplings in the skid piping. Because of these couplings, the pressure loads are unbalanced and must be reacted by bending of the pipe. The inspector found the licensee's calculations to be conservative and provided an adequate basis to support their conclusion that the "A" EDG weld failure was fatigue induced due to diesel vibration. The analysis for the reworked large bore piping (full penetration welds) showed that all stresses met code requirement The inspector reviewed the final safety analysis report and agreed with the licensee's determination that it contained no commitments regarding the design requirements of the diesel s%id mounted piping. Since this piping is mounted on a skid attached to the diesel, it is considered a part of a component, the diesel, and is not subject to piping design criteri The original purchase order for the diesels requested ANSI B31.1 criteria for the pipin However, the vendor took exception to this request, and the units were accepted with the piping built to manufacturers' standards. The manufacturer's drawings that were provided at the time of purchase showed full penetration welds for the large bore piping, even though the welds were not full penetration, and this design detail may have satisfied the purchasing agent's concer The inspector reviewed the vibration monitoring program initiated by the licensee. The program was found to be comprehensive and was implemented in accordance with the criteria and procedures described in ASME ISI Standard OM-S/G-1990. The inspector reviewed the acceptance criteria and concluded they were computed in a manner consistent with the standard. The inspector next requested the licensee to determine if the peak motions occurred in fundamental modes of vibration of the skid piping. This is a requirement if the basic criteria of the standard are used. This was corroborated and l
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further validated the acceptability of the developed acceptance criteria. The inspector j determined that the vibrations monitoring program for the EDGs was acceptable and should l provide early indications of system problem l
The inspector reviewed the Design Change Notices used to implement the modifications to the skid mounted piping and discussed these changes with the licensee's system enginee )
All partial penetration welds in the large bore piping were reworked to full penetration 1 welds. The new welds are non code welds and meet the requirements of the licensee's )
Welding Manual. In some instances, existing large bore tubing was changed to Schedule 10 pipe of comparable diameter and wall thickness Visual examinations of the reworked welds were made by the licensee in accordance with ANSI B31.1 criteria. The inspector observed some of the rework in process and found it was being performed in a professional manner. No modifications of the small bore piping and tubing were made as these were determined to be acceptable, as installed, based on walkdown inspections by experienced pipe stress engineer Conclusions The licensee's corrective actions were acceptable to address the August 2,1997, event in which the "A" EDG was rendered inoperable due to a vibration induced weld joint f ailure that resulted in the spill of 5 to 7 gallons of lubricating oil before the diesel was secure All welds in the large bore piping of both EDGs have been reworked to full penetration
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welds bringing the diesels to a condition comparable to the condition thought to exist at the time of initial purchase, and a vibration monitoring program was established to assure that exhibited vibrations meet the established lirnits. The licensee has shown that design satisfies the ANSI B31.1 code stress criteria. The NRC concluded that licensee's operability determination dated June 11,1998, that considered the diesels to be operable and fully qualified was acceptable. LER 50-336/97-30is close E8.4 MOV Corrective Actions Reaardina 2-CH-429 Inspection Scope (92903)
The inspectors reviewed NU's resolution of a recent condition report regarding a seat leakage problem for Unit 2 MOV 2-CH-429 which is the charging header isolation valve (2" Velan solid wedge gate valve). The review included the results of NU's root cause analysis of this problem to determine any impact on other valves at Units 2 and 3. The details of this issue are discussed in Millstone Unit 3 report details, paragraph E.1.1.2.
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i Roooet Details Summarv of Unit 3 Status
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Unit 3 was in hot shutdown (Mode 4) at the beginning of this inspection period. On April 30,1998, following the completion of work on the turbine driven auxiliary feedwater (TDAFW) pump, operators placed the unit in hot standby (Mode 3). Normal operating pressure and temperature were achieved in the RCS on May 1. After an unisolable leak on loop D isolation pressure relief valve (RCS*V132) was identified, the unit was cooled down to Mode 4 on May 8 and cold shutdown (Mode 5) on May 10 to facilitate repai (See Section M1.1 of this report for further discussion of this issue.)- Following the repair of RCS*V132, Mode 4 was again achieved on June 2. The operators subsequently heated the RCS to Mode 3 on June 9. On June 25, the plant was again placed in Mode 4 to
. facilitate the repair of a leaking auxiliary feedwater injection valve. Upon completion of this work, the plant was returned to Mode 3 on June 28, where it remained through the end of !
the inspection perio In a June 15,1998 letter, the NRC Commissioners voted to change Millstone 3 from a NRC watch list Category 3 plant to a Category 2 plant, effectively transferring approval for plant restart to the NRC Executive Director of Operations (EDO). This decision was based -
on many licensee, NRC Staff, and independent contractor briefings of the Commission, including the two final meetings conducted on May 1 and June 2,1998. Subsequently, on June 29,1998, the EDO granted final approval for the restart of Millstone Unit 3. The licensee commenced the restart on June 3 The Operational Safety Team Inspection (OSTI) was completed on May 5,1998 with a public exit meeting. The OSTI findings were documented in NRC inspection report (IR) 50-423/97-83. The results of a June followup inspection of the OSTIissues concerning operator performance and configuration control are documented in Sections 01.2 and 0 of this report. In addition to these reviews, the inspectors screened the operability determinations issued after the OSTIinspectio In the NRC Notice of Violation (NOV) and Proposed imposition of Civil Penalties issued to Millstone Station on December 10,1997, several escalated enforcement items (Eels)
documented in previous Millstone inspection reports were cited as NOVs and each were provided a NOV letter unique identifier. Closure of several of these items is discussed in the following sections of this inspection report. In previous NRC inspection reports, while these specific Eels were considered to be technically closed, the associated NOVs remained administratively open for further review, as required, of the licensee's response to the violations and the NRC determination as to whether further inspection of the corrective actions was warranted. As documented in Section 07.4 of this inspection report, all of the Unit 3 NOVs associated with the December 10,1997 Imposition of Civil Penalties have now been reviewed and are herein administratively closed.
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U3.1 Operations U301 Conduct of Operations 01.1 General Comments (71707)
The inspector conducted frequent reviews of ongoing plant activities, including observations of operator evolutions in the control room; main control board walkdowns; and inspections within the radiologically controlled area and other buildings housing safety-related equipment. In addition, several maintenance planning, plan-of-the-day, and plant
. operations review committee meetings were attende The inspector observed portions of various plant heatups and cooldowns. The operators generally conducted thorough briefings of upcoming evolutions, stressing the expected plant responte and potential problems and contingency actions. Operations supervision stressed there was no time pressure and the goal was safe, event-free operatio Operators performed these evolutions deliberately, in accordance with approved procedures, and using appropriate self-checking techniques. (An exception to these
- observations is documented in Section 01.4 of this report.) The inspector independently-verified selected prerequisites were completed before the evolutions were performe At the end of this report period, prior to the plant proceeding to Mode 2, the inspector identified the TS which would become effective in Modes 1 or 2 and verified that the licensee's reactor and plant startup procedures, ops 3202 and 3203, respectively, included the appropriate-TS in their Mode checkoff lists. In addition, the inspector reviewed surveillance test data for selected equipment to ensure the TS were met prior to entry into the applicable mode. All reviewed surveillantes were appropriately completed, as require The inspector also accompanied plant equipment operators on their rounds in the plan The inspector identified several minor seal, packing, and oil Isaks which had not had a trouble report issued for them. After the items were brought to the attention of the J
- operators the appropriate trouble reports (TRs) were written and associated work orders ,
' generated. Although this observation is indicative that more attention to detail may be l required by operators during their rounds, when the items were identified appropriate I actions were take .2 Ooerator Performance issues Inspection Scope (92901)
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The inspector reviewed the licensee's corrective actions in response to a restart issue concerning operational events identified in the Operational Safety Team Inspection (OSTI),
'NRC Inspection Report 50-423/97-83, dated June 12,1998. The team concluded that the formal root cause evaluations and corrective actions to address weak operator performance during the plant heatup should be completed prior to plant restart. In addition, the inspector reviewed corrective action documentation and conducted interviews with selected plant personne :
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i l Observations and Findinas l During the OSTIinspection, two operational events and three failures to meet technical
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specifications (TS) during the plant heatup were reviewed by the team. The events included inadvertent opening of a pressurizer power operated relief valve, an automatic initiation of the auxiliary feedwater system, not having the required number of operable reactor coolant system loops while transitioning into Mode 4, the failure to record the pressurizer temperature data during a plant heatup, and the failure to perform a surveillance I test when required. The OSTI team concluded that while there were no safety consequences as a result of these events, the performance by some plant operators during the plant heatup from Mode 5 to Mode 3 was weak. A Notice of Violation was issued with the OSTl report that included four technical specification violations associated with the five operational event The licensee initiated separate condition reports (CRs) for each of the events. Corrective action packages were also prepared for each of the respective CRs that included root cause investigations, where appropriate, and completed corrective action documentatio Specific actions were completed to address the individual events, as well as actions to-address the aggregate impact of the events. Additionally, a CR was initiated based on l preliminary findings from the individual investigations that showed indications of
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commonality between some of the events. The licensee determined that the positive j identification of any common factor in the events and implementation of corrective actions )
addressing them would be beneficialin preventing future events from occurring. To '
address the OSTI conclusion of weak performance by some plant operators, a set of operational performance improvements were initiated and documented in a memorandum from the Unit 3 Director to the Director of Performance Evaluation, Nuclear Oversigh The inspector reviewed the root cause investigations and the associated corrective actions for each of the events. Interviews were conducted with the Unit Director, operations staff, and nuclear oversight personnel. The licensee's response to these events was comprehensive and thorough. For example, a shift manager (SM) was taken off shift to help manage the power ascension activities until 100% steady-state power operations was achieved. The responsibilities for this SM included assuring: adequate staffing resources; adequate training was provided for the operating crews for specific startup activities; and I advanced planning was in place for activities / evaluations that were performed during power ascension. Individuals from the Regulatory Compliance group are now available via l beeper to provide support to the SM around the clock. The role of the shift technical i advisor was strengthened in that accountability for compliance with TS action statements l was assigned. The licensee assigned a senior manager to observe control room activities around-the-clock during heatups, reactor startups, and surveillance / testing activities. The expectations for the role of senior manager observers were clearly documented in a memorandum to the selected individuals that were fulfilling the position. An additional reactor operator and senior reactor operator were assigned to each shift during significant i evolutions, such as mode changes or plant startup/ shutdowns. Nuclear Oversight also provided a control room observer for around-the-clock coverage, The operational performance improvement plans included the following corrective actions I to address performance deficiencies in the Operations Department. " Peer Checking" was
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implemented for all operational procedures and surveillance, both in the control room, as well as in the plant. A review of scheduled activities through 100% power operations was performed to identify tl.ose evolutions requiring additional staffing, training, or pre-planning. A one day meeting was held offsite with the Unit Director, Nuclear Oversight representatives, and Operations Department management to define " Operational Focus" and to evaluate the department's performance against that definition. A continuous improvement culture was established throuph self-assessments and feedback from within the Operations Department and the Nuclear Oversight performance evaluation group. The Unit Director now meets with shift management and the Operations Manager at the end of 1 each shift rotation to evaluate the crews' performance from the previous week. Weekly meetirgs were established between the Unit Director, the Director of Performance Evaluation, the Manager of Performance Evaluation, and the Operations Manager to d;scuss operational focus and performance. Finally, the shift manager and unit supervisor conduct daily feedback interactions with the oversight shift observer. The inspector also noted that the Unit 1 Director was tasked with providing assistance to the Unit 3 Operations Department to facilitate the operational performance improvements. Conclusions The inspector reviewed the licensee's root cause evaluations and corrective actions for the operational events documented in the Operational Safety Team Inspection, NRC inspection Report 50-423/97-83, dated June 12,1998. The corrective actions were comprehensive and appropriately addressed the deficiencies identified in the root cause evaluations, as ,
well as the operator performance issues.
01.3 Valve and System Alianment Proaram
) Insoection Scope (92901)
The inspector examined the licensee actions performed in response to a restart issue identified in the Operational Safety Team Inspection (OSTI), NRC inspection Report 50-423/97-83,deted June 12,1998. The item involved the valve and system alignment program. The OSTI report stated that the issue required corrective action prior to restart and that the licensee had to demonstrate that the valve and system alignment program was effectively implemented. The inspector reviewed corrective action documentation, interviewed operators and management personnel, and examined equipment in the plant. Observations and Findinas Backaround The OSTI report noted that the licensee had a history of system alignment control problems and that, in 1997, Condition Report (CR) M3-97-0485 was written documenting 85 CRs wetten for valve, switch, and breaker misalignments. The OSTIinspection report identified l several additional equipment misalignments as well as procedure weaknesses and failures i to perform scheduled alignment activities. A Notice of Violation was issued uvith the OSTI l report for ineffective corrective action in this are Eauipment Alianment Methodoloav Weakness The inspector examined the methods used by the licensee to perform system alignment The inspector noted that the licensee generally had alignment checklists for fewer situations than are available at most other reactor plants. The licensee's alignments were, in the judgement of the inspector, difficult for operators to use and consequently, more prone to error. For instance, the licensee's program generally started with an alignment for what could be called a " cold iron" plant condition and then modified or added to the alignment by operating procedures that took the system through startup and eventually aligned the system for normal operation. The consequence of this method was that an cNrator, in assessing the plant configuration, had to review the system prestartup status, the procedures that were subsequently performed on the system, and then the operator had to integrate all of that information. The equipment alignment records in the operations filet did not show the operator how the system was aligned at the time of inquiry but rathe r how it was aligned prior to the performance of a number of operating procedure The , nspector noted that similar observations of the weak licensee methodology were mado by several of the licensee's oversight groups and industry peer advisors. In respt nse, licensee management stated that they were cor:sidering adopting the more
.comrr,on industry practice of using mode dependent alignments. The inspector noted that NRC did not have a policy on this matter but would continue to monitor the success of the licensee's methods based on performanc Corrective Actions Taken in response to the OSTI findings and as follow up to their own findings, the licensee took extensive actions to correct identified equipment alignment weaknesses. The inspector examined the corrective actions and observed that corrections to program procedures had been made to strengthen controls on tracking of alignment deviations, to provide instructions for entering rationale with any "not applicable" (NA) entries, to reinforce the requirement that only procedure changes should be used to change the required position shown on equipment alignment sheets, and to strengthen controls on throttle valve positic,n Additionally, the inspector observed that adequate remedial corrective action was taken for the failure to perform independent verification of safety related instruments and the failure to perform instrument valve alignments for root stop valves which had been identified in the OSTl report. Likewise, adequate remedial corrective action was taken for electrical breakers without component labels, and to correct breaker alignment nomenclatur Similarly, the inspector concluded that the licensee had taken adequate corrective action for cases of missing locking devices for main flow path valves and for the failure to perform the periodic (once-per-refueling-cycle) verification of locked valves.
I The OSTI repert and licensee's reports had both noted that the licensee did not have an effective way of tracking components which were not aligned at the time of the performance of the equipment line up. This had been a longstanding problem and was noted in previous licensee CRs. The licensee's corrective action was to issue a procedure change which required tagging the omitted component and logging it in the shift turnover lo L
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The OSTI report identified instances where three valves were found out of position. The inspector examined the licensee's investigative and corrective action and concluded it was adequate. In one case, the licensee reoriented the valve since it was subject to inadvertent operation through bumpin The inspector examined the issue idantified in the OSTI report regarding some breakers and switches which had been omitted from the alignment sheets for the diesel generators and <
quench spray system. The inspector concluded that the licensee had sufficient positioning {
controls when the operating procedures and the initial alignments were collectively )
considered. Mcwever, the inspector observed that in this case also, the licensee's I methodology for equipment alignments could be improved by adding more detail to the alignment sheet {
Additionally, the inspector examined the issue identified in the OSTI report regarding air start distributor filter drain valves on the diesel generator skids. The valves were not labeled, not shown on the P&lDs, and were not on the equipment alignment sheets. The inspector noted that there were other similar skid mounted valves such as the diesel cylinder test valves which were not labeled, not shown on the P&lDs, and were not on the equipment alignment sheets. The inspector found that the licensee controlled these valves in their maintenance procedures. This practice is not uncommon in industry. The licensee considered that any mispositionings would be evident on operator rounds. Further, the inspector noted that the vast majority of skid mounted valves were labeled and were J controlled by equipment alignment sheets or procedures. The inspector concluded that the l licensee's practices for skid mounted valves were adequat l In regards to the licensee demonstrating that the valve and system alignment program was effectively implemented, the inspector observed that 28 CRs (regarding equipment positioning related issues) were written in May and 31 CRs were written in the first % of June. This data suggested that the licensee's program was not functioning adequatel The inspector noted that a portion of the CRs were as a result the OSTIitself, a portion was due to the licensee's OSTl follow up actions, and some were due to an audit by the licensee's oversight group. However, a significarit number of the CRs were identified by problems found in the field, in order to assess whether they had achieved a level of performance sufficient for restart, the licensee identified a number of actions which were to be completed prior to restart. The actions included performing and assessing the results of a 5% sample of alignments for the month of June, performing a 100% active tagout review, performing a plant walkdown for unauthorized temporary modifications, and completing the immediate corrective actions for the Level 1 CRs issued as a result of an oversight audit completed on June 19,1998.
I Further, the licensee planned, starting in July 1998, to do additional special examinations of system alignment controls through monthly 5% samples of equipment alignments, monthly management observations, and followup audits and surveillance. The licensee considered that the data from that sampiing would enhance the ability to track performance in this area. The licensee also established a multi-disciplined, multi-unit Task Force for positioning issues to provide recommendations to management to improve the equipment positioning progra _
33 Conclusions The inspector examined the licensee's actions performed in response to a restart issue identified in the Operational Safety Team inspection concerning the valve and system alignment program. The inspector considered that the licensee had successfully demonstrated * .i the valve and system alignment program had been effectively implemente .4 Plant Heatuo and Transition to Mode 4 Insoection Scone (71707)
The inspector observed licensee performance in the Unit 3 control room during plant heatup and subsequent transition from Mode 5 to Mode 4, which was accomplished in accordance with OP 3201, " Plant Heatup." Observations and Findinas l
{
The licensee conducted a pre-job brief on June 2,1998, with the applicable operating crew i and discussed overoll procedure objectives, precautions, and basic procedure steps, as well as specific criteria necessary to administratively declare entry into Mode 4. In addition, the operating crew was briefed on SP 3601F.4,"RCS Pressure Isolation Valve Test," which was performed concurrently with OP 320 While the crew performed applicable steps within OP 3201 in support of the transition from Mode 5 to Mode 4, the inspector identified a procedural noncompliance that occurred l just prior to the declaration of Mode 4. During the heatup, a control operator was I monitoring the plant heatup rate in accordance with plant procedures, as well as peak reactor coolant system wide-range (WR) loop temperatures, and periodically reported these ,
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values to the unit supervisor (US). As WR temperature increased greater than 195'F, the operating crew failed to direct personnel to close and lock the applicable residual heat removal (RHR) outboard drain valve,3RHS*V26, as required by Step 4.2.34.g.2 of OP i 3201. The 195 *F wide range temperature limit provides margin such that a mode change into Mode 4 does not occur when temperature increases greater than 200*F without entry into a technical specification (TS) action statement for containment integrity due to the open RHR drain valve. Consequently, while a TS containment integrity violation did not occur, the established margin of the 195'F temperature limit was decreased based on the procedure noncompliance. The RHR drain valve 3RHS*V26 was subsequently closed as WR temperature approached 197'F during performance of SP 3601F.4. While the plant heatup cor'tinued after 3RHS*V26 was verified locked closed, Mode 4 was administratively declared when temperature increased greater than 200 * TS 6.8.1.a. requires, in part, that written procedures shall be implemented covering the activities referenced in Appendix A of Regulatory Guide 1.33 February 1978, item 2.a.,
General Plant Operating Procedures, " Cold Shutdown to Hot Standby." The licensee's failure to implement the plant heatup procedure as they transitioned frorn Mode 5 to Mode f
4, i.e., they failed to perform a required procedure step, constitutes a violation (VIO 50-l 423/98-208-04)of NRC requirements.
I m
l l 34 Conclusions The inspector concluded that while the licensee's performance during the plant heatup and transition from Mode 5 to Mode 4 was good, a violation of TS 6.8.1 was identified.
l Specifically, the licensee failed to properly implement the plant heatup procedure by not performing a procedure step that required an RHR drain valve to be closed and locked prior to exceeding 195' Fin the reactor coolant syste .5 Multiole Rod Droo Time Testina Insoection Scooe (61726. 71707)
The inspector observed portions of the pre-evolution briefing and performance of procedure SP 3451N22, Multiple Rod Drop Time Testing, on June 10 and 14, and reviewed the surveillance results to verify that the test was performed in accordance with procedures and the TS acceptance criteria were met, Observations and Findinas
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The rod drop test was performed as an infrequently performed test (IPTE) and as such had i a designated management test lead. A one-time change to the procedure designated it as i an IPTE to stress the importance of the test and focus plant personnel on the reactivity manipulation. Prior to commencing the testing, the operations crew trained on the procedure in the simulator. During this training and prior to test performance as the reactor operator was again reviewing the procedure, procedural issues were identified and corrected prior to test performance. The observed pre-evolution brief in the control room on June 10 was thorough and included the purpose of the test, procedural steps, and operational experience and acceptance criteria. Management stressed that there was no time pressure and encouraged control room and field personnel to voice any observed problems during the test since it had not been performed in over two year The inspector verified that selected prerequisites had been completed prior to the test and confirmed the acceptance criteria identified in the briefing were correct. Before the first rod was withdrawn on June 10, the test was terminated due to indications of high vibration on the secondary plant motor driven feedwater pump. On June 14, during the initial pull of the first rod bank the test was terminated due to a mismatch between the digital rod position indication and step counters. The inspector noted that the reactor i operator noted this deviation immediately and informed shift management prior to meeting l the mismatch criterion for terminating the procedur In each of the above instances, plant and operations management r%ponded appropriately to terminate the testing and place the plant in a stable condition. Because the rods were
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withdrawn in the second instance at the time of test termination, the reactor trip E-O !
procedure was entered and followed to restore the plant. The inspector observed professional and knowledgeable response by operations personnel during both of these instance ,
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i 35 The test was successfully performed on June 18. The inspector reviewed the completed surveillance test data and verified that all rod drop times were within the Technical l Specification 3/4.1.3.4 limit of 2.7 seconds and the surveillance acceptance limit of 2.19 i
seconds.
l Conclusions Rod drop time testing was well controlled and performed in accordance with approved i procedures. The management decision to conduct the test as an infrequently performed
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test focused personnel on the irnportance of the reactivity manipulation. Terminations of the procedure were appropriately performed and operators, using the surveillance test and emergency procedures, placed the plant in a safe condition following both secondary plant equipment problems and rod position indication discrepancies. The rod drop times were confirmed to be in accordance with plant technical specifications and procedural acceptance criteri U3 02 Operational Status of Facilities and Equipment O2.1 Charoino Pumos Mechanical Seal Leakaoe I Insoection Scope (71707)
On April 29, during a plant walkdown with a plant operator, the inspector observed boric acid accumulation at the "A" and "B" charging pumps' mechanical seal, in the seal basin, and the seal basin drain. In addition, the seal basin drain hole appeared to be completely blocked. The inspector discussed the deficiency with the operator, operations management, and the system engineer, and reviewed the impact of tne accumulated boric ;
acid and the engineering processes for evaluating and correcting the deficienc ' Observations and Findinas The system engineer confirmed that no trouble report existed for this condition, and that no activities were in place to monitor the leak or to clean the seal basin and the basin drain. The inspector determined that the operators and the system engineer demonstrated a lack of questioning attitude concerning the buildup of boric acid on the pump over an extended perio The inspector's concern was that, with the seal basin drain blocked, a potential exists for
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increased leak-off water to fill the basin and subsequent pump damage. The charging
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pumps are safety-related and as such, are required to operate during a design-basis acciden The system engineer determined that the existing seal leakage for both charging pumps l was normal (very small) for this type of seal, and that the seals are not designed to be zero leakage. No immediate safety issue exists with current plant conditions or with the blocked drain. However, a sharp increase in se< leakage rate will challenge this existing condition. The system engineer has taken corrective actions to ensure cleanup of the seal, basin and drain piping after the corresponding quarterly surveillance run of each pum I
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36 Conclusion The inspector identified that boron crystals had blocked the mechanical seal leakoff drains in two safety related charging pumps. The inspector also determined that the possible failure mode of increased leakage without drainage was not evaluated, and that no actions were in place to evaluate the mechanical sealleakage and to prevent blockage of the seal leakoff basin and corresponding drain line. The system engineer properly addressed the inspector's concerns, and initiated adequate corrective action .2 (Closed) Unresolved item URI 423/96-01-07: Safety Grade Cold Shutdown Eauioment Controls (Closed - SIL ltem 14) Insoection Scope (92901)
Following an update of this item in NRC IR 50-423/98-207two items remained ope These questions were associated with main steam atmospheric relief bypass (MSARBV)
block valve operability and cooldown and depressurization of a faulted steam generator j following a steam generator tube rupture event with a failure of the MSARBV. The (
inspector reviewed licensee actions to resolve these question I Observations and Findinas l l
The previous inspection identified a concern that the wording of Technical Specification !
3/4.7.1.6, associated with the MSARBV block valve, was ambiguous. The TS requires the valve to be open but does not specifically discuss whether it must be able to be closed to !
be considered operable. The licensee stated that the safety function of the black valves included the capability to be closed to isolate a stuck open downstream valve. Also, TS surveillance requirement 4.7.1.6.2 requires that the block valves be operated through a complete cycle to verify operability. The licensee's position was that if a block valve failed to cycle it would be declared inoperable and the appropriate TS action statements would be entered. This position was documented in operations memorandum MP3-O-98-023, dated March 31,1998, and has been distributed to all shift managers. The inspector concluded that the licensee has adequately addressed this concern and this item is closed, l
Another question identified in the previous inspection involved the ability to bring the plant I to residual heat removal (RHR) initiation conditions as described in the FSAR in the event of i a steam generator tube rupture (SGTR) event coincident with a failure of the MSARBV on the faulted steam generator. The licensee provided three procedures that are in place to perform a cooldown of a steam generator with a tube rupture. These are:
e EOP 35 ES-3.1, Post-SGTR Cooldown Usina Backfill:
e EOP 35 ES-3.2, Post-SGTR Cooldown Usina Blowdown: and, o EOP 35 ES-3.3, Po.st-SGTR Cooldown Usina Steam Dum The preferred method is to cool down using backfill since this minimizes radiological f releases from the faulted steam generator. This method allows the steam generator water to flow into the reactor coolant system through the break while adding feedwater to the steam generator to accomplish steam generator cooldown and depressurization. Steam is
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not intentionally released to the atmosphere from the faulted steam generator and the 1 MSARBV is not required to be operate j The cooldown using blowdown method drains the faulted steam generator through the blowdown lines and the addition of feedwater cools and depressurizes the steam generator. This method also minimizes radiological releases and does not rely on operation of the MSARBV All three procedures contain steps to monitor the boron concentration and to borate as necessary to maintain an adequate shutdown margin. The method using backfill requires sampling hourly since the backflow from the steam generator would dilute the reactor coolant and decrease boron concentratio The inspector reviewed the procedures and associated documentation and determined that they addressed the previous concern Conclusions The inspector concluded that the licensee adequately addressed the remaining questions that resulted from an inspection of SIL 14. SIL 14 is hereby close U3 03 Operations Procedures and Documentation O3.1 Procedure Uoarade Proaram Review (Closed - SIL ltem 80) ! Scope of the Inspection (92901)
The quality of and adherence to procedures had been a chronic problem at the Millstone site for all three units. The staff assessed the various programs and corrective actions that the licensee implemented to resolve this issue at Millstone Unit Observations and Findinos The issue of the need to improve procedure quality had been an element in the " Improving Station Performance" (circa 1995) program and the earlier " Performance Enhancement Program" (circa 1992). Both of these programs were on a Unit basis rather than on overall station program. In response to NRC concerns, the licensee developed the Procedure Upgrade Program (PUP) in 1992 to improve station procedure quality on a site wide basi The licensee's PUP commitment was included in a letter to the NRC dated June 4,1992, which described the licensee's overall Performance Enhancement Program. Because of the licensee's long outstanding commitment to complete ths PUP and past procedure adherence and quality problems, the satisfactory performance of the licensee's PUP was addressed as a separate issue in the NRC Restart Assessment Pla Although procedure improvement had been ongoing since the late 1980s, the licensee
, made a commitment to improve procedures to reflect industry standards for format and standardize procedures at all three units in the PUP. This commitment was made in the letter dated June 4,1992, as noted above. As a result of this process, the station l
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document control administrative proceduras were developed to be applicable to all three units. Recent inspections by the NRC have verified that most of the commitments made in the 1992 letter were met. As of May 1998, the Unit 3 PUP has been essentially 1 completed except for two procedure NRC reviews of the PUP program and status were documented in irs 50-423/97.-01,97- i 203 and 98-207. These inspections determined that Millstone Unit 3 had met most of its procedure upgrade commitments made to the NRC in the June 4,1992, letter. Three NRC ICAVP inspections of the licensee's configuration management program have been completed: 50-423/97-206,97-209 and 97-210. The three inspections reviewed a combined total of 97 licensee technical procedures with only minor violations identified. In addition, NRC OSTI Inspection 50-423/37-83,in part, reviewed procedure adequacy and !
implementation during licensee operations in Modes 4 and 3 (plant non critical heatup). i The OSTI concluded that the quality of operating procedures was generally good. With ;
few exceptions, the procedures reviewed by the OSTl were technically accurate and {
provided an appropriate level of detail. Also, recent licensee performance indicators by . !
both the licensee's Nuclear Oversight and their ime organintions indicate acceptable performance in both the areas of procedure performance and procedure technical adequac Conclusions The above NRC inspections and the licensee's own evaluations indicate that Unit 3 procedures are currently acceptable. SIL ltem 80 is therefore close U3 07 Quality Assurance in Operations 07.1 Review of NUREG-0733.TMI Action Plan Reguiremen J ji(92901)
(Closed - SIL ltem 38) ;
, Scope of th_e Inspection (929QU l As part of the licensee's review of the design and licensing bases, the NRC selected the programmatic area of NNECO's response to the Three Mile Island Action Plan, NUREG 0737 for detailed review of the implementation statu Observations and Findinas !
l The NRC originally reviewed the implementation status of the NUREG-0737 items in the l MP3 Safety Evaluation Report (SER),in its supplements (SSERs), and in inspection reports
! prior to and subsequent to the issuance of the original license in January,1986. In irs 423/97-207 and 208, the NRC reviewed the current implementation status of certain NUREG-0737 items. Based upon the findings, the NRC requested that the licensee perform i
additional reviews of the NUREG-0737 items, including: items that had findings by the NRC in these reports, items with outstanding CRs cgainst them, and items not yet reviewed by NRC. The licensee performed these additional reviews and submitted the results to the NRC in letter B17239, dated May 11,1998. This letter provided the status ,
of all the items reviewed. The inspector reviewed the letter and noted that it addressed !
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the NRC's request for additionallicensee reviews of compliance status. Thire were several TMI items noted in the letter with outstanding compliance issues identified. These had Action Request assignments to establish compliance prior to startu The inspector noted that a substantial binder of information had been preparet to document the compliance bas!s for each item covered by the letter. The inspector also selected two items and binders to review: 1.D.1, Control Room Design Reviews, and ll.K.1.10, Operability Status. The licensee concluded that they were in compliance 'with these two items, and the binder provided substantial information to substantiate this finding. The inspector concurred but noted that there were outstanding ARs related to item 1.D.1 with activities scheduled in the future. These actions were deemed to be acceptable by the inspecto The inspector also reviewed the updated FSAR Table 1.10-1, titled TMl Action items, which incorporated changes from reviews conducted during this outage in order to make the Table an accurate representation of the current implementation status for Unit 3. The inspector noted that a substantialimprovement had been made in the Tablo and noted only a few minor administrative items that needed correction. The licensea initiated action to correct these items. These actions were also deemed acceptable by the inspecto In a letter (B17306)to the NRC dated June 11,1998, the licensee clarified the status of remaining actions required to restore full compliance with NUREG-0737 commitment One of these action items, regarding the post-accident sampling system, is discussed
- further and closed as part of Sllitem 83 in Section P8.1 of this inspection report. Other action items included pl.:nt license amendment requests, which have been submitted to the NRC Office of Nuclear Rector Regulation (NRR) for review. Inspectors have reviewed, as applicable, operability determinations related to the completion of the NUREG-0737 items, noting that, where required, the emergency plan and operating procedures have been revised to reflect current plant configuration and design basis assumptions and analysi Conclusions The inspector has determined with reasonable assurance that Unit 3 is in substantial compliance with TMI Action Plan requirements. SIL ltem 38 is hereby close .2 Final Safety Analysisfeoort Chance Process (Closed - Sll,ltem 2) Insoection Spffe (40500)
In IR 50-423/96-201 an NRC Special Inspechon Team reviewed Nuclear Group Procedure (NGP) 4.03, " Changes and Updates to Final Safety Analysis Reports for Operating Nuclear Power Plants," which was the primary procedure to ensure that the Final Safety Analysis Report (FSAR)is properly maintained. The team noted severa; instances in which the l
Millstone Unit 3 FSAR was inconsistent with other licensing-and design-bases documents, procedures, operating practices, or the as-built plant configurations. The six deficiencies noted :n which the Millstone Unit 3 FSAR was not maintained current or did not reflect the actual plant configuration or operating practices coactituted an apparent violation of 10 CFR 50.71(c) (eel 50-423/96 201-01). Significant item List (SIL) Item #2, "FSAr' 'pdate"
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was included on the Millstone Unit 3 SIL to review the licensee's corrective actions for the apoarent violation and to evaluate the licensee's overall FSAR change process. The FSAR change piocess was reviewed by several NRC groups, including the headquarters project {
manger, the resident inspectors, NRC Independent Corrective Action Verification Program (ICAVP) inspection teams, and the ICAVP independent contractor, Sargent and Lund This inspection included a review of the NRC inspections from January 1,1997, to the present (including the inspection reports generated by the NRC ICAVP team inspections)
and the ICAVP independent contractor (Sargent and Lundy) interim report, which was issued on May 8,1998, as they related to the licensee's FSAR change process. The inspection also reviewed the licersee's March 2,1998, response to the NRC Notice of
. Violation and Proposed imposition of Civil Penalties, which was issued on December 10, 199 The inspector noted that nn December 18,1997, the licencee superseded NGP 4.03 with Regulatory Affairs and Compliance Procedure (RAC) 03, " Changes and Revisions to Final Safety Analy:is Report." Both procedures were reviewed during NRC inspections and the following summary documents the staff's overall review in this are Observa.tions and FindiDgs The licensee's corrective actions for the apparent violation (eel 50-423/96-201-01)
identified in IR 50-423/96-201 were reviewed in IR 50-423/98 206. As discussed in the inspection report, the inspectors concluded that the licensee's corrective actions adequately addressed the technical aspects of the issues and the individualitems were considered c!osed. However, the Notice of Violation (NOV - letter unique identifiers 01262,01272,01282,01292,01302, and 01312) for each issue would remain administratively open. The inspectors also reviewed NGP 4.03, and RAC 03 which superseded i:, and stated that the procedure adequately defines the guidance to evaluate whether a se fety evaluation is needed and the areas it must addres As documemed in irs 50-423/97-207and 50-423/98-206,the inspectors randomly selected and Javiewed approximately 20 FSAR changes which were submitted to the NRC in letters dated July 1, August 23, October 29, and November .21,1997. The NRC staff reviewed the FS AR Change Requests (FSARCRs) to ensure that the proper safety evaluations (10 CFR 50.59 safety evaluation) were completed and appropriate, that the licensee adequately addressed 10 CFR 50.59 for the intent changes to the FSAR, and that the classification of non-intent changes were appropriate. The NRC staff determined that each specific intent change was properly addressed h a 10 CFR 50.59 safety evaluation and, therefore, no unreviewed safety questions were identifie >
The NRC IC.AVP out-of-scope inspection (50-423/97-206) identified two instances where the information contained in the FSAR had not been updated within the past 24 month However, the team did not evaluate the overall FSAR change proces The NRC ICAVP Tier 2/ Tier 3 inspection addressed the FSAR change process in IR 50-423/97-209. The ICAVP teem reviewed NGP 4.03 and RAC 03, as we!! ns a number of FSARCRs. Many of the FSARCRs reviewed by the team were, in fact, multiple changes I combined in one package. In accordance with the licensee's NGP 4.03 procedure, those
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l changes which were classified as clarifications rather than changes did not have a safety
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evaluation completed prior to approval. Within the group of changes which were for clarification only, the team identified some changes that affected the licensing basis of Millstone Unit 3. As such, the team concluded that those changes snould have included a safety evaluatior-The team also concluded that the threshold for writing a safety evaluation was not
, appropriate. This observation was discussed with the licensee and the licensee indicated that both internal and external reviews in June 1997 identified this threshold program weakness and corrective actions had been initiated. The corrective actions included (1)
- reviewing all the FSAR changes performed from the t!me of licensing until the start of the l
Millstone Unit 3 configuration management effort which was in 1998, and (2) revising the applicable plant procedures and re-training plant personnel concerning the development of formal safety evaluations for FSAR changes which were previously described as clarifications. The team further identified an issue with the licensee's corrective actions in that the corrective actions did not included a review of approximately 500 FSARCRs approved between the start of the configuration management program and the retraining of personnelin 1997. However, the team found that those FSAR changes implemented since mid-1997 appropriately include a safety evaluation when needed. Furthermore, the team -
l tound that those approved since the fall of 1997 were more complete and included discussions of all the changes included in the FSAR change package. The team attributed these improvements to the corrective actiors described by the licensee, clear goals and l expectations from the ooerations review comm'ttee, and critical self-assessments in this
! area. The team did not have any concerns with the new guidance for FSAR chanDes contained in RAC 03, or the cernpleted FSARCRs.
i During the implementation of the ICAVP, Sargent end Lundy reviewed the licensee's processes for making changes to the Millstone Unit 3 licensing and design basis. In the l interim report on the ICAVP, Sargent and Lundy stated that the design of plant modifications implemented after receipt of the plant's operating licerise was technically
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adequate and configuration control was adequately maintained. Sargent and Lt.pdy further stated that the licensee has established programs, prncesses, and procedures to maintain
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effective configuration control of their licensing and design basis in the future.
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The inspector reviewed the licensee's Reply to a Notice of Violation and Proposed i imposition of Civil Penalties, dated March 2,1998, as it related to the FSAR change l process. The corrective actions listed were reviewed during the inspections addressed abava. As such, the NRC staff finds the licensee's corrective actions acceptabl _C_onti usion_g Based cn reviews ar,d inspections noted above, the inspector concluded that the licensee's FSAR change process is adequate and should maintain effective configuration control of the Millstone Unit 3 licensing and design bases. Although some issues with the implementing procedure were identified (most with the older versions of the procedure),
the NRC concluded that the licensee's corrective actions and revised procedure RAC 03, adequately control changes to the fac!!ity. Therefore, the inspector concluded that SIL ltem 2 is hereby closed. The NRC Notice of Violation (NOV -letter unique identifiers l
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01262,01272,01282,01292,01302, and 01312) included in the NRC letter dated December 10,1997,is also close .3 9_ggective A.ction Proamm (40500_) (Closed - SIL ltem 37)
Backaround i
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The NRC has been dealing with Northeast Nuclear Energy Company (NNECO) on broader )
performance issues, which go beyond compliance with the design and licensing bases concerns. These broader concerns are considered contributory causes for the past poor ,
performance. The datient concerns included leadership, communications (employee concernt), the corrective action program, procedural adherence and procedure upgrades, work planning and control, and operational enhancement The licensee's corrective action program was weak in the identification of problems and ineffective in ensuring comprehensive and effective corrective actions. There were many (
instances of narrowly focused corrective actions that failed to embrace the causes of the l
underlying problem. Additionally, the licensee failed to follow up on corrective actions to !
ensure they were effective. Also, the NRC believed that the immense number of I allegations received from the Millstone station were directly related to the inability of f allegers to get known problems corrected through the formal corrective action progra Consequently, the RAP determined that the licensee's corrective action program was a restart issu Licensee Actions The licensee initiated efforts to improve the corrective action program by adopting industry standards and prccesses and formalizing them in procedure RP 4, " Corrective Action Program." It is a site-wide program and has been in effect since February 1997. The fundamental changes to the process involved the lowering of the threshold level for ]
reportable problems, management emphasizing the need for employees to identify j problems, greater management involvement in the process, timely processing of operability determinations, development of performance indicators, training in root cause analysis, and enhancement of the tracking and trending programs. Improvements to the process have resulted in about six revisions to the governing procedure, RP 4. In addition, NNECO has established tracking systems for corrective actions that are not identified by a CR, e.g.,
operating experience, training, preventive maintenance, employee concerns, and engineerin Management established a corrective action assessment program and performance indicators to monitor the corrective action program effectiveness. The Nuclear Oversight l
Organization developed its own program for assessing the quality of the corrective action L program and provided NNECO management with independent evaluations of several L attributes of the corrective action process. NRC management closely monitored these performance indicators throughout the recovery process and discussed the justification for any reported successes with NNECO management at periodic public meeting __
NRC Inspection Activities Because of the large number of Condition Reports (CRs) [ Note: CRs were previously called Adverse Condition Reports) being identified by the licensee's staff during the recovery process, the resident and regionalinspection staff concentrated on specific issues identified by the CR process and audited the licensee's corrective actions for completeness. The staff periodically selected additional CRs for review, based on the licensee's assigned level of importance, or their risk significance, as perceived by the resident staff. Additionally, other CRs were examined to provide a spectrum of safety significant and lesser risk issues. The intent was to primarily assess the corrective action program while dealing with the safety significant technical issue The NRC Restart Assessment Panelidentified licensee activities that would be evaluated to obtain an integrated assessment of the licensee's corrective action program. For example, inspection activities were focused on corrective actions for enforcement items, the Significant items List, the restart deferred items, employee concerns, independent Corrective Action Program (ICAVP) findings, self-assessments, commitments, licensee event reports and degraded and non-conforming conditions. Examination of the corrective action program also included the review of the Action Requests (AR) from the Action item Tracking and Trending System (AITTS) program, which is an extension of the CR proces Improvements in the corrective action process were judged by the completeness of the licensee's corrective actions for each of the inspected areas within the corrective action program. The inspection process examined the identification and processing of problems, the assessment and root cause evaluation, the directed corrective actions and the closure J
of the issu A significant input to assessing the licensee's corrective action program was derived from the normalinspection program where valuable insights regarding the effectiveness of ;
corrective actions are routinely collected from the technical safety inspections. Also, major contributions to the measure of the effectiveness of the NNECO corrective action program for Unit 3 were the NRC inspections performed from February 9 through 20,1998 using procedure 40500," Effectiveness of Licensee Controls in Identifying, Resolving and Preventing Problems"; on April 13,1998, the NRC started the last of the NRC Independent Corrective Action Verification Oversight Branch assessments of the licensee's corrective actions for degraded and non-conforming conditions; and from April 13 through May 5, an Operational Safety Team inspection (OSTI), which audited portions of the corrective action process during the course of its activitie In addition to direct inspections of the NNECO corrective action program, the NRC performed an inspection, IR 50-245,336,423/97-212,of the employee concerns program
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and safety conscious work environment at the Millstone station in conjunction with the Order issued on October 24,1996. As discussed previously, the NRC believes that the
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ineffective corrective action program spawned many of the allegations brought to the NR The report essentially found substantial improvements in the safety conscious work environment. This finding supports the IP 40500 inspection findings of improved management / worker communications and the workers being able to freely identify safety issues with the expectation that they will be properly addresse )
44 1 Routine insoection Findinas The NRC inspection staff selected identified problems that were being processed within the licensee's corrective action program and entered them into the NRC Significant Iterns List, item 37, that is part of the NRC Restart Assessment Plan. These items represented safety-related technical issues, in most cases identified by NRC inspection such as violations, that l required closure by the licensee. Additionally, assessments were made of the licensee's
! corrective action program as part of the routine resident inspection process. The resolution ( of the technicalissues allowed the NRC staff to evaluate the technical adequacy of the
! corrective actions while assessing the effectiveness of the licensee's corrective action l
program. The staff found that, as time progressed from the initiation of the licensee's new corrective action program, the quality of the closure packages improved. This improvement was noted in NRC irs 50/423-97-02;97-202; 97-203; and 97-20 Additionally, NRC team inspection report.s using inspection procedure (IP) 40500 (IR 50-423/97-82)and the OSTI (IR 50-423/97-83),as well as the NRC corrective action review of the independent corrective action verification program (ICAVP) results (IR 50-423/98-211), all indicate appropriate reviews were conducted and the licensee's progress in corrective action program implementation was acceptably demonstrated.
l Conclusions The NRC staff examined the licensee's corrective action program for the identification and processing of conditions adverse to quality, the assessment and root cause evaluation of those conditions, and the identification and implementation of the corrective action They also observed licensee management involvement in the process to ascertain the proper application of appropriate expectations, standards and overall support to the progra It is evident from the inspection record and the licensee's performance indicators that an i
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appropriately low threshold exists at Millstone Unit 3 for the identification of adverse l conditions to quality. There has been an order of magnitude increase in the number of I Condition Reports written since the implementation of the new corrective action program l and the raising of management expectations and standards. Licensee management has effectively communicated these new standards to the working staff such that problems are l identified and placed in the corrective action program for resolution. Additionally, the overall process is capable of producing the required result if it is properly implemente l l
As a result of the numerous inspections of the Corrective Action Progrem, including team i inspection results from the IP 40500 and OSTI inspections, the implementation of l corrective actions at Unit 3 are deemed to be adequate and SIL ltem 37 is hereby close J 07.4 Significant items List (SIL) Evaluation and Issue Closure (92901)
(Closed - SIL ltems 20,43,68,71, & 72)
l Certain SIL items represented programmatic issues that were assessed under the purview l
of contractor activities and verified by NRC overview. One such program was the Employee Concerns Program (ECP), which was monitored by Little Harbor Consultants, In l
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(LHC), and independently evaluated by the NRC, along with the safety-conscious work environment (SCWE) at Millstone Station. NRC inspection efforts of both the ECP and SCWE and the evaluation of LHC's oversight activities are documented in irs 50-L 245,336,423/97-212and 98-210, both issued in April 1998. In the former report, licenst,a corrective actions to address a violat:on (96-59-13)of 10 CFR 50.7a were reviewed and found adequate. Overall, the findings and conclusions docurnented in these l two inspection reports constitute the substance of the assessments necessarv to address the ECP/SCWE issues and concerns. Therefore, SIL ltem 20 has been appmpriately reviewed and is hereby close Sirnilarly, NRC assessments of the ICAVP, including the evaluation of the independent l contractor (Sargent & Lundy) activities, have been in progress, as documented in several l NRC inspection reports issued over the past year. On June 17,1998, the Director of the Office of NRR iss ued a letter to the licensee indicating that the ICAVP has been completed to the satisfaction of the NRC, as required by Section IV.1 of the August 14,1996, l Confirmatory Order for Millstone Unit 3. Therefore, the SIL item addressing the ICAVP I Order, SIL ltem 43, is hereby close Three other Sllitems documented the NRC plan to conduct a review of all open inspection items to determine which, if any, issues required further followup and evaluation prior to the NRC staff consideration of a Unit 3 restart recommendation to the Commission, items encompassed by this review included allegations received by the NRC; Licensee Event Reports (LERs), which are routinely screened and evaluated by NRC inspectors for corrective action implementation by the licensee; and any enforcement issues and unresolved items, which require NRC inspection closur On December 10,1997, the NRC issued a Notice of Violation (NOV) and Proposed imposition of Civil Penalti6s (CP) to the licensee for a number of violations applicable to the three Millctene units, including 29 findinge previously documented as escalated enforcement items (Eels) for Unit 3 alone. Over the course of the last 18 months, as documented in several Unit 3 inspection reports, the technical aspects of these Eels have been inspected and closed, with the associated NOVs remaining administratively ope With the issuance of this current inspection report, all the Unit 3 Eels associated with the December 10,1937 Civil Penalty package have now been reviewed and close Therefore, all the NOVs related to the 29 Eels (each provided with its own five-digit CP letter unique identifier) are also hereby administratively close Specific unresolved items were reviewed and closed in this current inspection report, as part of the NRC review of restart issues. Likewise, allissued LERs were screened and several selected for detailed inspection followup, as is documented in the previous two and this current inspection report. Finally, while allegations are not normally documented as such in rottine NRC inspection reports, those allegations applichble to Millstone Unit 3
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and/or Mi!! stone Station programmatic activities were specifically reviewed for safety l issues that could impact the restart of the unit; and were appropriately dispositioned.
l Therefore, the three SIL items covering the NRC review of open restart issues described
, above (i.e., SIL ltems 68,71 & 72) are hereby considered closed.
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l U3.ll Maintenance U3 M1 Conduct of Maintenance M1.1 Evaluation of Temocrarv Modification of a Reactor Coolant System Valve Insoection Scop _q1D2707)
The intent of the inspection was to provide NRC inspection oversight of an emerging, critical maintenance activity using the attributes of Inspection Procedure 6270 Observations and Findinat l
With the plant at normal operating pressure and temperature in Mode 3, the licensee identified a small packing leak on valve 3RCS*V132, a 1 %" globe valve in a line cross-connecting the reactor coolant system (RCS) loop "D" isolation valves. This valve is normally open to provide a pressure relief path from a filled loop with the reactor coolant isolation vaives closed. The licensee cooled and depressurized the RCS to Mode 5 (i.e.,
- < 2OO'F) and attempted two repacking repairs; the first with RCS pressure approximately
! at 350 psig and the second with the RCS depressurized. Both repairs were unsuccessful
- because of disc / stem separation, preventing the valve from being "back-seated". The l second repair attempt resulted in an unisolable RCS leak of approximately 3 gpm, which I was subsequently reduced to below 1 gpm with the installation of a clarnp by maintenance personnel.
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The licensee established both an event review tearn (ERT) and an independent review team ( (IRT) to investigate and review the sequence of events, decision points, and causes related to the V132 packing leak, repair attempts, and resultant unisolable RCS lea Subsequently, after the installation of a " freeze seal" to isolate the valve from the RCS, the valve packing was replaced to stop the leak and the valve reassembled without repair to l the damaged disc / stem assembly. A temporary modification (3-98-028) was installed to allow movement of the disc off the valve seat, while remaining engaged with the stem, and to provide the pressure relief function for any trapped fluid in the 1 %" diameter piping
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upstream of V132. The temporary modification also provided for the installation of a stem
- ejection stop plate, as a replacement for the valve " backseat", as the back-up function to l the stem thread engagement for missile protection. The installation of this temporary l modification allows the existing valve V132 configuration to remain in place until the next L refueling outage when valve repair / replacement can be effected at reduced RCS water l inventor The inspector reviewed the final reports provided by both the ERT and IRT investigations, examined the safety evaluation and design details associated with temporary modification 3-98-028, observed maintenance activities related to the set up and use of a V132 mock-up assembly to troubleshoot and provide training for the freeze seat implementation and subsequent valve repair / modification work, and viewed the radiographs confirming V132 disc / stern separation, along with certain engineering details discussed in the temporary modification package. The inspector also reviewed technical evaluations for the V132 lift
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pressure (M3-EV-980122) and the V132 disc behavior during the current operating cycle (M3-EV-980133);and discussed with the cognizant design engineering personnel the code compliance and functional criteria associated with the approval of temporary modification 3-98-028. Givan the function of this valve, the inspector raised two safety-related issues relative to the modified V132 installation; i.e., restoration of the reactor coolant pressure ,
boundary integrity and prevention of loose parts being introduced into the RCS. Based '
upon the review of the relevant design documents and interviews with engineering personnel, the inspector concluded that operation of the plant with the temporary modification installed is in compliance with regulatory requirements, meets the intant of documented regulatory commitments, and is deemed a safe approach to addressing the existing valve condition. The following points address and document the inspector's conclusions in this regard:
- The primary function of this valve (to prevent over-pressurization of the "D" reactor coolant loop, when isolated) is not required during normal operation, since operation of the plant with a loop isolated is not permitted in operational modes above cold shutdown (Mode 5) conditmn * The valve integrity, as part of the reactor coolant pressure boundary, was restore l to its original design configuration and no valve leakage is currently evident. (The !
licensee has monitored this through the end of this inspection period and will ;
continue visual inspection of the "D" loop area, using a camera during reactor coolant system heat-up and subsequent power ascension.)
- Installation of this temporary modification provides sufficient missile protection, as l backup to the primary valve design feature precluding stem ejection; thus, meeting i the intent of FSAR commitment j i
- Radiography of the valve internals has confirmed no evident loose parts, apart from 1 the stem / disc separation. The licensee safety evaluation appropriately evaluated the l impact of flow and vibratory conditions on the valve internals, concluding the installed configuration prevents loose parts from rnigrating into the reactor coolant loop o This temporary modification was reviewed for compliance with ASME Code requirements and Generic Letter 96-06 guidance. Operating conditions were confirmed to be maintained within ASME Code limits for the structural integrity of the pressure boundary and with sufficient margin maintained for over-pressure ,
protectio l
- The modified configuration of 3RCS*V132is considered capable of adequately maintaining the reactor coolant pressure boundary and performing any required over-pressure protection safety functio Conclusions j l i
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Based upon the review of the relevant design documents and interviews with engineering j personnel, the inspector concluded that operation of the plant with the temporary l l
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modification installed is in compliance with regulatory requirements, meets the intent of documented regulatory commitments, and is deemed a safe approach to addressing the existing valve conditio The inspector noted that both the ERT and IRT reports provide recommendations intended to address problems identified with the valve 3RCS*V132 repair activitiss, The inspector confirmed the license's intent to repair or replace ihis valve, restoring the modified valve to its original design configuration, during the next Unit 3 refueling outage. The inspector has no further questions regarding the status, functionality, or future plans regarding 3RCS*V13 U3 M3 Maintenance Procedures and Documentation M3.1 Material, Eouioment, and Parts Lists (MEPL) Proaram Review (Closed - SIL ltem 25) (Closed) URI 50-423/95-07-10: Unit 3 MEPL Proaram implementation issue _g Inspection Scope (92902)
The Unit 3 MEPL program was reviewed in irs 423/97-202,97-203 and 98--207. IR 423/98-207 updated the open issues and closed a number of them; this report provides a further update and resolutio Observations and Findinas PMMS Identification Numbers: The numbering scheme for PMMS results in differences i between the identification for components in the field /on drawings and in the PMMS database, when the number of characters exceeds 15. The licensee performed a review of the PMMS database and found that there were 392 Category 1 identification numbers that I have been reduced from their full set of characters . They further reviewed and categorized each item as to what was removed to reach the 15 character limit. The !
licensee developed and issued a new specification, Specification SP-M3-ME-027, Rev. O, l
" Standard Specification for Processing of Technical Data in the PMMS ID BOM Database,"
dated March 12,1998. This specification provides guidance for data conventions, including character reduction, when required. Most of the truncations made to the numbers were trivial and not of concern, but there were about ten that had the asterisk *
dropped (in accordance with the new Spec. SP-M3-ME-027), creating the potential for mistaking a safety related item for nonsafety related . The licensee wrote CR M3-98-2343 on 5,'5/98 to address this issue. Corrective action # 2 for this CR includes a modification to the specification to address this concern. The due date for this action is December 10, 1998. This will be reviewed by the NRC as itern 1 of IFl 423/98-208-0 ]
Designation of Safety Related components with an asterisk (*): FSAR, Figur_a 3.2.2 stated that an usterisk indicates that equipment is quality assurance Category 1 (i.e., Safety l Related). However, the inspector noted that: 1) not all SR components use the asterisk as noted in the FSAR, e.g., SR snubbers; 2) there is some ambiguity in the use of the asterisk for relays; 3) and some identification tags and signs in plant do not use the asterisk, even though the component is SR and the asterisk is used in PMMS. The licensee noted that in l
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49 l accordance with the plant's c,riginalidentification convention there are a small number of types of components that may be SR, yet not use the asterisk in their identification .
The licensee issued FSARCR 98-MP3-2 to revise the FSAR to accurately reflect the actual !
plant practice as defined in their specifications. The licensee also issued Specification SP- ]
M3-ME-024, Rev. O, Conventions for System identification, System Interfaces, and j Equipment identification, that currently defines in detail the conventions used to establish i and maintain system and equipment identification for MP3 systems, structures, and I components. The inspector toured the plant and observed a number of components and their labeling, then compared these labels to that in the PMMS and MEPL systems. The inspector noted that not all cumponent identification conformed to the new Specificatio The licensee's label gruup corrected some of these items. The licensee also wrote CR M3- l 98-2343 on 5/5/98 to acdress this issue. Corrective action # 1 for this, CR includes a modification to the specification to provide guidance for equipment identification labels for {
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" Relays, Snubbers, Junction Boxes, and Appendix R equipment." The due date for this j action is December 3,1998. The inspector discussed with the licensee whether the in-plant labels would then be corrected after the Specification was revised, and the licensee revised the CR corrective action to address this question in December after the l Specification is revised. This will be reviewed by the NRC as item 2 of IFA 423/98-208-0 {
The PMMS database for Unit 3 was noted to be incomplete. Some safety related (SR)
components are not in the MP3 database, e.g., snubbers. Many augmented Quality Assurance (QA) and nonsafety related (NSR) parts and components are also not in the database. MP3 has assignments (ARs) to add the snubbers to the MP3 database and to f add all augmented QA components by December 1998. The licensee has stated that they will meet this date and have the PMMS database updated to contain all MP3 safety related items (including snubbers) and all augmented quality items by the end of 199 A related issue was noted in the response to the Notice of Violation (NOV) associated with eel 96-201-43 (Civil Penalty unique identifier 04043)that states that one hundred percent of the Unit 3 components in the PMMS database were re-evaluated for the proper quality category. However, in order to adequately address some of the MEPL concerns, it is important to re-evaluate all components (in the safety related and augmented quality categories) for the proper quality consistent with plant design and licensing bases. The inspector discussed with the licensee whether those items not in the PMMS database had been reviewed and, if not, when they would be. The licensee stated that they had not beenreviewed, and that the statement in the NOV response should be interpreted as applying to the MP3 PMMS database. However, the licensee also stated that these items would be re-evaluated as part of the effort scheduled to be completed by December. 199 The licensee also stated that the NOV response wou!d be updated to include these additional corrective actions. This will be followed up by the NRC as item 3 of IFl 423/98-208-0 Previous problems were noted with the MEPL evaluations and PMMS entries associated with the Unit 3 containment hatch in Unresolved item 423/95-07-10. These were updated in irs 423/97-202 and 423/97-208. Since that time, the licensee has performed additional work on the MEPLs, bill of materials (BOMs). PMMS, and drawings associated with the hatch. As a result of these efforts, the licensee has divided the hatch into 50 separate ,
local component identifications: 43 of these 50 were newly created daring the current l
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effort. Thkty-nine of these 50 components have BOMs with a total of about 1000 part ' Twenty-one new MEPL evaluations have been performed (9 at the component level and 12 at the BOM level). Sixteen components were upgraded from NSR to SR as part of these efforts. There were six CRs and six NCRs written to disposition discrepancies and upgrade The inspector reviewed a sample of the newly generated CRs, NCRs, MEPLs, drawings, and PMMS involved with the containment hri;h, and inspected the hatch and related components in the plant. Discussions were held with the cognizant engineers. With the exception of the following two items no further discrepancies were found.
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NCR 398-005, covering the upgrade of four pressure equalizing valves on the hatet, was dispositioned as "use-as-is" without adequate justification. The licensee investigated the '
parts replaced on the four valves and determined that the material was acceptabir,, but that documentation was not complete, thus a new CR, M3 98-2771 was inued. ThL4 will
, ensure that documentation is completed to verify that appropriate materials are installed in I the valves. The second item was that a drawing change to properly identify hatch interlock limit switches had not been completed. The licensee issued a CR change form to i M3-98-1321 to add corrective action #5 that would address the drawing. This is --
(- acceptable and unresolved item URI 50-423/95-07-101s closed.
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! M3.2. Review of eel 96-201-4: Land ACR M3-96-0912:Ingdeauste closure of MEf1 l related NCRs I
ACR M3-96-0912 addresses Escalated Enforcement item 423/96-201-43. These items f were closed in IR 423/98 207. The NRC issued a Notice of Violation and Proposed
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imposition of Civil Penalties by letter dated December 10,1997 that includes this item with
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NOV letter unique identifier 04043. NU subsequently responded to this NOV with letter
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B16996, dated March 2,1998. Page 129 of this response letter addresses NOV 0404 The inspector reviewed the NOV response and noted that allitems were addressed by the licensee actions and NRC reviews performed in IR 423/98-207,with the exception of the l additional items discussed in the MEPL sections herein. The response notes that each unit
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will be in full compliance prior to restart. Therefore, the response to NOV 04043 will be closed and the completion of the other items herein will be separately tracke .
M3.3. Issue of Non-Safety Related(NSR) Parts in Safety Related (SRI Components l The issue of NSR parts in SR components has been raised by previous NRC inspections and by a number of internal CRs, engineering reviews, and oversight reviews. This area was also updated in NRC IR 423/98-207,but some areas remained to be resolved. These remaining areas are discussed her The licensee issued two Technical Evaluations, which summarized the various activities taken to address this concern: Technical Evaluation for Acceptability of Installed Parts and Materials in Maintained QA Category I plant Components, MP3, M3-EV-98-0022, Rev. 2, May 12,1998 (TE-0022); and Technical Evaluation for Work History Review for i Components identified in NCR 397-010, Rev.1, dated May 12,199 .
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The inspector had previously questioned some of the numbers of components documented in Rev.1 of TE-0022. Rev. 2 of TE-OO22 has clarified the numbers that fall into the various categories of components that received different types of reviews for the NSR part concern. However, the inspector noted that one of the categories of components (containing 402 components) in Rev. 2 of TE-0022 had used inappropriate criteria for determining acceptability. That is, the licensee called replacement parts acceptable if they used the " original design basis component criteria." This is the same problem that was identified in IR 423/98-207 with respect to NCR 397-010,that resulted in NSR parts being incalled into SR components. Following up on thb, the licensee noted that the category of 1430 components from TE-OO22 also partially used a similar criteria. The licensee, therefore, re-evaluated these two categories of component This new review resulted in the re-evaluation of the work history and parts installation history of 1496 components. The reviews determined that there were documentation or qualification questions about 52 components. CR M3-98-2667 was issued on May 27, 1998, addressing these 52 c'mponents. Thirteen components were judged to require only a MEPL evaluation, since they currently had none, but should be designated NSR. Thirty-nine parts (1'7 mechanical and 22 electrical /l&C) will require a successful commercial grade dedication process or replacement to return the component to a fully qualified status. In .
the interim, the licensee has completed an Operability Determination (OD MP3-070-98) and found all components fully operable. The inspector reviewed a sample of the work history review packages, the summary report, the CR, the OD, and back-up information. The inspector discussed with the licenses the basis for the OD for seven components; and the licensee provided adequate backup justification and documentation for their conclusion The schedule for returning all components to a fully qualified status is defined in the corrective action plan for CR M3-98-2667 as 8/15/98. This is acceptable and will be followed up by the NRC as item 4 of IFl 423/98-208 0 As a result of the reviews performed, about 100 components have been identified with parts installed in SR components that are not fully qualified as SR. These consist of: 3 from the original TE-0022 Rev. O review, 29 from the NCR 397-010 re-evaluation, 52 from CR M3-98-2667,and between 10 and 20 from ongoing maintenance work performed over i the last two years of the shutdown. CRs were written to address each of these identified !
items. The licensee plans to issue a new Technical Evaluation within the next few weeks !
to replace the two earlier'ones and to consolidate all of the findings. This is acceptable j and will be followed up by the NRC as item 5 of IFl 423/98-208-0 ;
M3.4. Current Parts Control and Consumable issues
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l M3. Parts Control '
The parts issuance portion of the program was not reviewed with the rest of the MEPL program in 1R 423/97-202 due to ongoing issues identified there by the licensee. The licensee implemented various preventive actions through 1997 to address this concer This has resulted in increasing controls over time. Nonetheless, some individual problems continued to occur. Some reviews showed good performance (e.g., Oversight )
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memorandum of Jan. 10,1998 - sample of 13 work orders with no problems; Oversight i l
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memorandum of Dec. 29,1997 - sample of 24 work orders with no problems), yet other reviews identified some ongoing problem areas (e.g.., CRs M3-98-0407 and M3-98 0989).
CRs M3-98-0407 and M3-98-0989 documented concerns with the procedures and processes for the control of parts and materials to be used during work in the plant. The lack of overall coordination and control of the process was also identified. As a result, a comprehensive corrective action plan was developed. One key element of the plan was that the Director of Materials Management was designated as the overall process owne Another key action was the establishment of a new group (3CONFIG) within the Unit 3 design engineering department whose function is to review and approve all material requests (MRs) associated with work on safety-related systems and components. The review and approval process of the MRs includes the following:
- verification of the accuracy of the MR source documentation,
- initiation of a Document Cliange Notice (DCN) to correct any documentation or drawing errors,
- verification that a MEPL is referenced when required or initiate a MEPL evaluation, ,
- verification that the part number on the MR agrees with source documents, I
- initiation of any necessary database update requests, and;
- verification that the quality level of the part is correc This group was established as a compensatory measure to ensure the proper parts end materials are used in the plant. In addition to the 3CONFIG group, a materials team, i designated 3PTIA, was also established to review and approve the MRs. This group will ensure that PMMS bill of material (BOM) is updated, that the Material Information ;
Management System (MIMS) database information is correct and that the part in the '
warehouse matches PMMS and MIM l
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The corrective action plan for CR M3-98-0407is comprehensive and proposes a number of procedure and process changes as well as the need for training of the personnelinvolved in the process. The implementation of the appropriate corrective action plan items is intended to result in the functions of the 3CONFIG and 3PTIA groups eventually being partially or fully assumed by various department personnel normally involved in the process. At the time of this inspection most of the corrective actions had not been completed, with many scheduled to be completed in the next several month M3. Consumables During IR 423/97-203,the inspector noted an issue with control of consumables, as follows. There was not, at that time, an effective method for tracking MEPL status on consumable items, or for ensuring that any requirements established in the MEPLs were met during the purchasing and use of the item. The Hard Copy MEPL, which is used for i i items that did not have a specific component identification (such as consumables,) had not been kept current. The licensee stated that the MEPL program would be updated to appropriately address the consumable items. They also stated that the information from the MEPL evaluations, the chemical consumable product list (CCPL), and from Procurement would be properly integrated, and that the MIMS system would be updated to track any
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MEPL requirements for consumables to ensure that they are addressed when purchasing l
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the consumables, in 1998, the licensee performed a further review in this area, identified ongoing consumable problems and issued CR M3-98-040 The licensee has taken actions to improve the control of consumable items and the corrective action plan for CR M3-98-0407 contains several additional actions that are intended to improve the control of chemical consumables. These include ensuring that MEPL evaluations have been performed for consumables, and development of a consumables list for use by work planning and maintenance personnel. The licensee has l
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issued procedure MC 6, " Quality Assurance for Nonsafety items," which provides directions for procurement, inspection and analysis of nonsafety chemicals for impuritie The licensee was arranging for chemical analysis of a sample of chemical consumables in stock to confirm that the level of any impurities was within acceptable limits. Chemical consumables issued from the warehouse that have the potential for use in a safety-related application are reviewed by the 3CONFIG group. This will aid in ensuring appropriate controls, including any associated MEPL evaluations, are implemented. Also, work planning procedure WC-1 has been revised to add instructions to ensure that parts and consumables are recorded on the work orde The inspector concluded that the licensee has recognized the need to improve the control-of parts and consumables and has taken actions to implement improvements. Also, the corrective action plan for CR M3-98-0407is comprehensive and many of the items were in j the process of being implemented. Nuclear oversight has been involved and continues to l closely monitor resolution of the issues. Additional NRC follow up will be performed, as j item 6 of IFI 50-423/98-208-05,on the implementation of the remaining portions of the )
corrective action plan and a review of the effectiveness of recent program improvements, '
M3.5. NSR Parts Uoarade As part of an extensive evaluation, the licensee discovered 29 components on NCR 397-010, that had NSR parts installed in the SR components. The licensee wrote CRs and performed Operability Determinations (ODs) for each of them. The ODs concluded that the components were still operable. In general these parts were procured from the original )
equipment rnanufacturer to the original specification. Also, the licen.see was able to verify the following types of information for many of the 29 components (but not all of these ;
aspects for every item): procurement documents / purchase requisitions, some QA review l of AWO, proper part stock code numbers, materialissue forms (MiFs) with proper number, drawings, maintenance performed to procedures, vendor calls, successful post maintenance testing, and successful periodic surveillance testin The inspector previously requested of the licensee a plan and schedule for bringing the 29 l degraded / nonconforming items into full compliance with Appendix B requirements, as i discussed in Generic Letter 91-16, Rev.1. The licensee completed corrective action plans
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(CAPS) for each of the three CRs written to address the 29 items. These CAPS specify a
! program to research and identify specific part level information and to conduct a I commercial grade dedication effort in order to return the components to a " fully qualified"
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status. These activities are scheduled to be completed by August 16,1998. This will be l followed up by the NRC as another part of item 4 of IFl 423/98-208-0 ,
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54 Conclusions The Unit 3 material, equipment, and parts lists (MEPL) program was reviewed in several inspections over the past year. The licensee has invested substantial effort into improving the program and has significantly upgraded both the program and the evaluations for many cornponents and parts in Unit 3. A number of issues were identified during the review and the licensee has been responsive in addressing the concerns. The program is currently deemed acceptable and meets regulatory requirements. The licensee is continuing to upgrade the program and the NRC intends to follow six aspects of these activities as inspector follow item IFl 423/98-208-05,as noted above. ACR M3-96-0912 was inspected and eel 96-201-43 of SIL 25 was closed in inspection report 423/98-207. The technical aspects of the NOV with unique letter identifier 04043 were reviewed and closed and URI 95-07-10is closed herein. Additionally, the various open issues identified during previous NRC inspection reviews of the MEPL program are also closed herein. Therefore, SIL ltem 25 is close U3 M8 Miscellaneous Maintenance issues M 8.1 Control and Use of Vendor Information ACR 10562 Review (Closed - SIL ltem 18) Scope of the insoection (92902)
The overall site and Unit 3 program for Vendor Interface and vendor manuals were reviewed in irs 423/97-203 and 98-207. The various aspects of the program, as outlined in NRC Generic Letter 90-03 and the licensee's procedure DC 16, were satisfactorily reviewed with the exception of the final review and updating of procedures to address l vendor manual changes made during the vendor manual update process. That aspect is updated here, Observations and Findinas I
The inspector selected and reviewed two vendor technical manuals (VTMs) that had been updated per the new DC 16 process, namely: Manual 25212-004-001, Rev.1, installation, Operation and Maintenance of Service Water Pumps, and Manual 25212-001 -035, Rev.1, installation, Operation and Maintenance of Reactor Trip Switchgear. The inspector reviewed: the controlled manuals held in the several nuclear document libraries; the procedure reference sheets that had been completed per DC 16 by Mechanical Maintenance, instrumentation & Electrical, Operations, and Generatinn Test Services; and I the procedures that were referenced as pertaining to the manuals. The appropriate incorporation of selected vendor information into the procedures for the service water pumps and reactor trip breakers was also reviewed at the same time. The inspector also discussed the process and documents with personnelin the various organizations that l
were involved in the manual update and procedure review process and observed the pumps and circuit breakers in the plant. With the exception of the below items no discrepancies were identified:
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55 The control and filing of the procedure reference sheets was noted to be weak. CR M3-98-2505 was issued by the license . Not all of the pertinent procedures appeared to be listed, e.g., OP 3326 for the SW Pumps. The inspector also discussed with the licensee how the review for operating procedures was performed, and what has been accomplished to date. CR M3-98-2505 was issued by the licensee? There were differences in the recommendations of the VTM ano OP 3326 for the startup and shutdown of the SW Pumps. The licensee stated that the VTM in this area is somewhat generic to Hayward Tyler pumps, despite the fact that it is a manual for the NU contract. Also, the MP3 plant-specific design with an automatic pump discharge valve takes precedence over the generic information in the VT This was deemed acceptable by the inspecto CR M3-98-2505 specifies corrective actions to: reconcile and correct VTM procedure reference sheets in controlled copies of VTMS, assure the correctness of operations procedure references on the Service Water Pump manual and other approved VTMs, and revise DC 16 to improve the handling of procedure reference sheets and the process associated with operations procedures. At the end of the inspection period, Rev. 2 to DC; 16 was in the review process. Corrective actions were confirmed to have been completed i or were in an acceptable state of progress.
i f Conclusions The corrective measures associated with the vendor information problems documented in ACR 10562 have been addressed, as inspected and verified during several report period Therefore, SIL ltem 18 is hereby close M8.2 (Closed) LER 50-423/97-010-00: Electrical Calculation Discrepancies in Minimum
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Voltaae Analysis for Class 1E Electrical Systems
! Insoection Scope (92700)
In January 1997, with Unit 3 in Mode 5, a NRC review of electrical calculations associated
! with Class 1E,480V and 120 V systems identified discrepancies and non-conservative l assumptions between related electrical calculations used to demonstrate design basis compliance. These discrepancies included the use of assumed voltages which were greater than those which had previously been calculated to exist under degraded voltage conditions. These elevated voltages were used as input to calculations performed to evaluate the 480V and 120V systems. These analytical-limit, worst-case minimum values were not utilized at the source voltage for separate voltage-drop calculations i performed for the 480V and 120V bus loads. Degraded-grid-voltage relays are required to ensure that safety related equipment and devices either have adequate voltage to perform their safety functions or are not damaged due to a degraded voltage condition.
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o 56 Observations and Findinas The inspector reviewed LER 97-010-00,the " Supporting Design Change Pcckage Details",
the "Walkdown Checklists", the " Engineering Release Transrnittals", the " Turnover l transmittals" along with other associated documentation and engineering materials
' showing that this work was completed. The licensee reviewed the associated calculations, the set points for electrical components and electrical component performance parameters.
, As a result of this review the licensee made extensive changes in the electrical supply equipment and the associated calculations. The changes made included making a number of transformer voltage tap changes to higher voltages, replacing relays, adding equipment, rnaking under-voltage relay set point changes, installation of new 480V system over 1 voltage alarms and other miscellaneous circuit modifications for specific loads to ensure
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acceptable minimum terminal voltages on other electrical equipmen I Conclusions The completed corrective actions taken by the licensee are deemed adequate. Therefore, LER 97-010-00is close ].
M8.3 (Closed) LER 50-423/97-011-00: Hvdroaen Reg _ombiner Heaters Potentially Outside i of Desian Basis Under Dearaded Voltaae Conditio Insoection Scope (92700)
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The analyticallimit, worst-case minimum voltage values were not utilized as the source l voltage for separate voltage drop calculations performed for the 480V and 120V bus loads.
j If the 4160V bus voltage was at its analytical limit worst-case minimum value, individual l devices supplied by the distribution system could have inadequate voltage to perform their l design basis function. Therefore, the ability of tia Hydrogen Recombiner heaters to perform their safety related function was questioned. The licensee committed to "perforrn
, a calculation or evaluation to evaluate the Hydrogen Recombiner heater performance under )
l degraded voltage conditions" and to " implement design changes to restore the Hydrogen '
! Recombiners to their design basis, if required." Observations and Findinas The inspector reviewed CR M3-97-0161, along with other associated documentation, calculations and engineering drawings. The licensee commissioned the vendor of the Hydrogen Recombiners (Rocketdyne Boeing) to evaluate their product's performance at !
degraded voltage. The design basis of the recombiners is to maintain hydrogen l concentration below 4.0 volume percent. Rocketdyne Boeing determined that the minimum power requirement was 25.8 kW for the Hydrogen Recombiner to function and j completely combine all hydrogen at maximum flow rate and at a 0.5% by volume of I hydrogen. Further eva!uations established the heater performance at the design basis, degraded plant voltage of 87.16% or 413.6 volts to be 26.73 kW. The heater output of i 26.73 kW at the degraded plant voltage is above the 25.8 kW require ]
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The Hydrogen Recombiner has 15 heaters. Random heater element failure was considered to occur at a rate of 0.42 elements a year, and was also analyzed. With one heater element failure, the power output, at the 86.17% degraded voltage condition, would be 24.72 kW. For complete hydrogen combination at maximum flow rate and 24.72 kW, hydrogen concentration must be at least 0.7% by volume. The hydrogen concentration 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a LOCA, when Hydrogen Recombiner operation is initiated has been estimated to be in excess of 0.7%. Thus, the recombiners will satisfactorily function in this case. Even with up to three heaters failed, the recombiner will be able to complete the recombination reaction, at the degraded voltage. in this case, higher hydrogen concentrations guarantee hydrogen removal due to the additional heat of reaction. For three heaters failed,1.7% of hydrogen will react to completion. Thus, the design guarantees that the hydrogen concentration is maintained below 4%, as per the design basis. This is acceptabl Conclusions The hydrogen recombiners are operable at the postulated degraded voltage and are able to meet their design basis requirements. The actions taken by the licensee are adequat LER 97 011-00is close M8.4 LQlosed) LER 50-423/97-061-00: Technical Specification 6.8.4 Leakane Reduction l
Broarem Does Not Address all Leakaae Paths Irispection Scope (92700)
In December 1997, an NRC inspection Team for Unit 3 determined that there were leakage '
paths, during a post-accident situation, in systems outside the containment to the Refueling Water Storage Tank (RWST), which had not been previously considered by the i licensee. This was in violation of Technical Specification (TS) 6.8.4.a. These leakage paths are important because the fluids leaking can be potentially radioactive and the RWST l is vented to atmosphere. The leakage of radioactive fluid to the RWST provides a potential i source of radionuclides to the atmosphere not previously considered by the licensee in the offsite and control room dose calculations. The leakage of concern is the fluid which is able to leak through valves internal to the piping system and not visible as external leakage. To evaluate and reduce the potential leakage the licensee committed to: 1.) Perform a review j of systems outside of the containment that could leak highly radioactive fluids during a I serious transient or accident, to ensure potential offsite dose consequences are within the licensing basis; 2) Perform a leak test of each potentialleakage path to establish valve l baseline leakage, and incorporate the results into the offsite dose assessment calculation; l and 3) An overall MP3 leakage reduction program document will be prepared and program )
owner identified. Program elements determined to be inadequate will be addressed in j accordance with the Millstone Corrective Action Progra l l Observations and Findinas i i The inspector reviewed LER 97-061-00 along with other associated documentation and i engineering materials. The inspector reviewed the 18 corrective actions taken by the l licensee. The principle corrective actions taken by the licensee were; 1) the licensee leak i
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tested or evaluated 32 valves and heat exchanger.;, for internal and externalleakage summing up the collective leakage; 2) the licensee evaluated the feasibility and then determined that installing a HEPA filter on the vent of the RWST was not required; 3) the licensee reviewed and revised as necessary leak testing procedures to assure that leak testing in the future will be accomplished according to the requirements of TS 6.8.4; and finally 4) the licensee evaluated "The Post-LOCA Radiological Impact of Sump Coolant Backflow into the Refueling Water Storage Tank".
The licensee determined that the radiologicalimpact of a release from the RWST was similar to that of a Steam Generator Tube Rupture where releases of radionuclides are made through the Safety Relief Valve with the er.ception that the release through the RWST is only by the partitioning of the radionuclides from the stored water. No actual flashing of radionuclides in the RWST is anticipated. The computer code used in the Steam Generator Tube Rupture evaluation was used to evaluate the partitioning of lodine and the resulting whole body, skin and thyroid dose to personnelin the Low Population Zone and in the Main Control Room. The maximum 30 day dose was to the thyroid for both populations and when compared to the 10 CFR Part 100 limits were less than 1 % of the 300 Rem limit for the Low Population zone and Less than 3% of the 30 Rem limit for the Control room personne Conclusions The completed corrective actions taken by the licensee are deemed adequate. LER 97-061-00is close .111 Enaineerina U3 E1 Conduct of Engineering E Motor-Operated Valve Proaram Review (Tl 2515/109) (Closed - SIL ltem 26)
(Open) IFl 98-208-06: Comotetion of MOV Proaram Closure issues Backaround The most recent inspection of motor-operated valve (MOV) activities at Millstone occurred in February 1998 as documented in NRC Inspection Report (IR) 50-423/98-82. Open items were identified therein requiring resolution prior to closeout of the NRC review of Generic Letter (GL) 89-10, " Safety Related Motor-Operated Valve Testing and Surveillance" at Millstone Unit 3 as follows: one violation regvding MOV calculations, one unresolved item (URl) regarding a training documentation issue, and nine inspector followup items (IFis)
regarding various technicalissues. At the end of the inspection, the licensee issued a letter dated May 14,1998, stating that Millstone Unit 3 had fully addressed the program requirements of GL 89-10.
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E1. Resolution of MOV Proaram TechnicalIssues Insoection Scooe Northeast Utilities (NU) responded to the open items of IR 50-423/98-82in their letter dated April 25,1998. The inspectors reviewed the NU responses to the open items such as the " alternate test plan" (ATP) changes to the MOV program, and the corrective actions implemented and planned regarding the MOV thrust calculation errors. The purpose of the inspection was to complete the NRC review of GL 89-10 at Millstone Unit The review of the open items where certain corrective actions were incomplete is included below. This part of the review included four icts. Also, an issue concerning a seat leakage problem for a Millstone Unit 2 valve (2-CH-429, Charging Header isolation) that developed after issuing IR 50-423/98-82is discussed herein since it had MOV program implication The inspectors completed the remaining parts of the review, which included the violation regarding the thrust calculation errors, one URI regarding a training documentation problem, and five IFls, and determined that these items were complete with no need for further review. These items are addressed in Sections E8.1 - E8.7 of this repor i Observations and Findinas (Closed) IFl 98-82-01: Dynamically test 3 MSS *MOV 17A/B/D and address steam blowdown conditions for these and 3 MSS *MOV 74A/B/C/D valve I Valves 3 MSS *MOV74A/B/C/D are 8" Pacific globe valves which serve as the Steam Generator pressure relief bypass valves. In this application, these valves would operate under high temperature compressible fluid conditions. As noted in the NRC Safety
- Evaluation (SE) of EPRI's Topical Report TR-103237,"EPRI MOV Performance Prediction Program," the globe valve modelis applicable to cold-water (less than 150 F) pumped-l flow ccnditions. Therefore, NU was unable to apply the Performance Prediction Model
_ (PPM) to these valves. .At the time of IR 50-423/98-82,the ATPs indicated that valves 3 MSS'MOV17A/B/D (steam isolation valves to the AFW Turbine Driven Pump) would be dynamically tested and the results applied to 3 MSS *MOV74A/B/C/D to resolve the high
! temperature and compressible fluid concerns. The licensee committed to perform the
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dynamic tests 013 MSS *MOV 17A/B/D prior to restart of Unit 3. The inspectors noted that the proposed alternative testing does not address glob 3 valve performance under steam l blowdown conditions. Licensee personnel agreed to revise the ATPs to resolve the blowdown concer In its letter, dated April 25,1998, NU stated that the testing of 3 MSS'MOV 17A/B/D could not address the blowdown performance conditions for 3 MSS *MOV 74A/B/C/ ]
Therefore, the licensee was removing the reference to 3 MSS"MOV 17A/B/D testing for !
application to the MSS 74 valves. NU committed to reevaluate the ability to test l 3 MSS *MOV 74A/B/C/D under blowdown conditions and consider possible offsite testing j of a prototype valve. Commitment B17178-06was established to be implemented by the l licensee's action tracking and trending system (AITTS) number AR 980090420 Specifically, at least 2 MSS 74 vclves will eitner be differential pressure (DP) tested in situ
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or an offsite prototype valve test at design basis (DB) conditions will be conducted to l evaluate blowdown effects due to high temperature, compressible fluid flow condition The inspectors considered the licensee's revised position and commitment regarding these valves to be reasonabl l l
The inspectors noted that, in general, the DP testing of the 3 MSS 17 valves revealed l thrust requirements for high temperature steam conditions to be much greater than I predicted. A CR was issued to address the thrust underprediction. In accordance with MOV program requirements, NU was evaluating other globe valves for possible impact to ensure that appropriate area terms and high temperature / compressible flow effects were considered in MOV setup. The MSS 17 valves will be included as ATP valves because of j the high thrust requirements. Also, the MSS 74 valves will be considered low thrust margin valves in light of the MSS 17 valve DP testing result (Closed) IFl 98-82-02: Determine by test aoorooriate friction coefficients for stainless steel auide surfaces > 100 i NU has applied EPRI PPM results to several gate valves which have stainless steel guide l l
and slot surfaces that would experience fluid temperatures above 100 F which exceeds l the restrictions of the PPM model. The licensee has used friction coefficients for carbon steel-on-carbon steel surfaces as best available data. In its letter, dated April 25,1998, {
NU committed to sponsor with EPRI a test program that will determine the appropriate friction coefficients to use for stainless-on-stainless guide surfaces for fluid temperatures above 100*F. The specific NU commitment (B17178-07)will be implemented by AITTS AR9800904202 to be done prior to restart from refueling outage (RFO) LClosed) IFl 98-82-06: Clarify how test data will be obtained, evaluated, and documented to ensure that the assumed bearina coefficients for 3SWP'MOV102A/B/C/D remain adeauat NU has implemented the PPM to establish the torque requirements for Millstone's butterfly valves. Millstone Unit 3 has 4 Contromatics butterfly valves with bronze bearings that operate in raw water conditions (Service Water). The PPM would normally default to a bearing coefficient of 0.60 for these valves. However, the licensee had performed dynamic testing of these valves that supported use of a 0.20 bearing coefficien Therefore, this lower bearing coefficient was used in the PPM to establish the Contromatics' torque requirement During the conduct of Inspection 50-423/98-82,the inspector noted that NU needed to clarify how it would ensure that the assumed bearing coefficient for SWP"MOV 102A/B/C/D remained adequate. In NU's letter, dated April 25,1998, the licensee committed to perform future dynamic tests of these MOVs at the next RFO with possible
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additional tests. The licensee also committed to revise the ATP to discuss bearing l coefficient of friction degradatio During this inspection, the inspector determined that the bearing coefficients for these butterfly valves have been included in the ATP calculation. The commitment (B17178-11)
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to monitor the valve bearing coefficients during dynamic testing prior to restart from RFO 6 will be implemented by AITTS AR 98009042 06.
(Closed) IFl 98-82-07: Verify actions to be taken, includina modifications to 3SIL*MV8804A/B and 3SlH'MV8802A/B.to imorove the thrust marain for low marain valves.
NU did not have a formal margin improvement plan for low margin MOVs. This decision was based on the assumed conservatism of NU's setup methodology. However, the inspectors noted that several MOVs are justified in Alternate Test Plans using alternatives to NU's standard program methods. A risk ranking review of the ATP valves found that 4 MOVs,3SIL*MV8804A/B(high risk) and 3SlH*MV8802A/B(medium risk) had safety function margins of less than 6%.
In a letter, dated April 25,1998, NU committed to upgrade the capability of these MOVs before restart from RFO 6. This commitment (B17178-12)will be implemented by AITTS AR 9800904207. Conclusions The inspector verified the adequacy of NU's corrective actions as described in commitments regarding the technical issues included in IFis 50-423/98-82-01,-02,-06, and -07. These IFis are considered closed and new IFl 50-423/98-208-06is opened to track completion of the future actions regarding these MOV program issues. Based upon -
this inspection review and previous inspection results, SIL ltem 26 is hereby considered to )
be closed.
E1. MOV Corrective Actions Reaardina 2-CH-429 Inspection Scoce (92903)
The inspectors reviewed NU's resolution of a recent condition report regarding a seat leakage problem for Unit 2 MOV 2-CH-429 which is the charging header isolation valve (2" i Velan solid wedge gate valve). The review included the results of NU's root cause analysis of this problem to determine any impact on other valves at Units 2 and 3. . Observations and Findinas NU issued CR M2-98-1173 and 1177 on April 27,1998, to effect corrective actions regarding the specific problems with MOV 2-CH-429, including any MOV program implications at Millstone Units 2 and 3. The specific leakage problems with 2-CH-429 were corrected adequately. More importantly, a root cause analysis of the problem was completed on May 14,1998, where NU concluded that there were 3 contributing factors that caused the seat leakage problem: (1) Marginal bluing acceptance of the seating surfaces; (2) Additional static seating force; and (3) A variation of pressure pinching.
These factors combined to cause the valve disc to travel beyond the seating surface, resulting in the seat leakage conditio i I
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In addition to the specific repair of 2-CH-429, NU evaluated 12 other gate valves at Unit 3 '
(most of which were solid wedge gate valves also) and found that none of these valves were susceptible to the seat leakage problem of 2-CH-429. Also, no other similar '
problems were identified with other Unit 2 valves. These valves were eliminated from concern predominantly due to the fact that their static seating forces had not been significantly increased as was the case for 2-CH-429. Another corrective action taken by the licensee included a change to the valve maintenance procedure to provide quantitative tolerances for gate valves to assure that the lapping of disc / seats is done such that contact and bluing are well onto the seating surfaces and not at the edge Conclusions The inspectors concluded that the licensee's corrective actions regarding the seat leakage problems of 2-CH-429 were adequate and appropriat U3 E2 Engineering Support of Facilities and Equipment E (Closed) Walworth Valve Yoke Crackino. IN 93-97 (Closed - SIL ltem 51) , Insoection. Scoce (92903)
In NRC IR 50-423/98-82,an inspector documented the review of the licensee's action in response to NRC Information Notice (IN) 93-97," Failures of Yokes installed on Walworth Gate and Globe Valves," concerning cracking problems experienced with these valves at other nuclear facilities. During the current inspection, the inspector reviewed the status of .
DCR MP3-96-080," Generic Motor Operated Valve (MOV) Yoke Replacement," and the adequacy of NU Report No. M3-ERP-960002," Northeast Utilities Root Cause Investigation Millstone Unit 3 'Walworth/Aloyco Yoke Cracking Root Cause Analysis,'" dated February f
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2,199 Observations and Findinos The inspection included a review of DCR MP3-96-080, Rev. 00, dated 02/14/97. As noted in the DCR, in response to NRC IN 93-97, visual inspection of all Walworth MOVs {
was performed, cor.sistent with the inspection methods used by the manufacturer '
(Walworth). Following identification of a crack indication in MOV 3CNM-MOV79A, a complete inspection of yokes on MOVs manufactured by Walworth was performed using enhanced inspection methods of liquid dye penetrant testing (PT) and magnetic particle (MT) examination, as opposed to the visual inspection methods (VT) used by the l manufacturer. The disposition of any as-found flaws was then based on visual acceptance standards.
!
! Of the 78 MOVs inspected by the licensee,38 were found to have crack-like indication Of the 38 valves with as-found indications,15 of the valves are in the GL 89-10 MOV program, and 5 additional valves are designated safety related, but not part of the GL 89-10 program. The remaining 18 valves are designated as non-safety relate j
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During this inspection, the status of the yoke replacement activities for the 20 safety-related valves was reviewed. According to DCR MP3 96-080,the licensee decided to replace the yokes on 12 safety-related valves and 5 nonsafety-related valves found to have crack-like indications. The yokes on these valves were to be replaced with redesigned yokes. Of the 12 valves identified in DCR MP3 96-080 as requiring yoke replacement, Engineering Release Transmittal Form l's had been issued for all valves with the following exceptions: For 3RCS*MOV8000A/B,the RCS Pressurizer PORV Block Valves, a DCR Change Notice was issued indicating the valves had been released on 04/15/98. In actuality, the valves themselves, and therefore the yokes, were replace / For 3 MSS *MOV17A/B/D,all three valves had Engineering Release Transmittal Form I completed indicating that the valves had the yokes replaced as of 04/22/98, with !
the exception that 3 MSS *MOV17D had failed the post maintenance testin l J
Of the 5 non safety-related valves requiring yoke replacement, two valves, 3ESS*MOV478/C,did cot have Form l's but instead had Design Change Notices indicating 1 that the modification had been field verified. The remaining 8 safety-related valves and 13 '
non safety-related valves were repaired under Nonconformance Reports (NCRs) by welding and/or grinding under separate work orders not related to DCR MP3 96-080. Appropriate documentation was provided indicating completion of the work on these valves (included in Altran Corporation Project Report Nos. 96157,96163,96175, " Millstone Point Unit 3 -
Walworth Yoke Cracking Incident, Vol. 4).
The inspector reviewed NU Memorandum MP3-DE-97-0270," Distribution of Walworth Yoke Cracking Root Cause Evaluation per AR 96006255-04," dated 03/20/97,which attached NU Report No. M3-ERP-960002," Northeast Utilities Root Cause Investigation Millstone Unit 3 'Walworth/Aloyco Yoke Cracking Root Cause Analysis,'" dated 02/13/97. The inspector found the root cause analysis to be complete and comprehensive with appropriate analysis and photographs of as-found flaws in the samples examine The root cause analysis was distributed, as per the memorandum, to NU personnel at the j MP1 and MP2 units and to the Haddam Neck (Connecticut Yankee) and Seabrook facilities l and to the Millstone MOV Program Manage Therefore, the portion of the licensee's corrective actions in response to SIL 51 concerning Walworth valve yc,ke cracking as identified in NRC IN 93-97 is considered close l Conclusiqn_s in response to NRC IN 93-97 concerning cracking of yokes in motor-operated valves manufactured by the Walworth company, the licensee performed an evaluation of all 78 affected valves and performed a root cause analysis to determine the cause of the cracking observed in the yokes of several sample valves. The licensee repaired or replaced the yokes for all 20 safety-related and 18 non safety-related valves which had shown indications of cracking in the valve yokes. Therefore, the licensee's corrective actions in response to the concerns documented in IN 93-97 and Walworth valve yoke cracking have been deemed appropriate. SIL ltem 51 is hereby close l
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i 64 E2.2 Hiah Enerav Line Break (HELB) Review for the Auxiliary Feedwater (AFW) Svstem Pioina Inspection Scoce (37551)
l The past closure of the Turbine Driven Auxiliary Feedwater (TDAFW) pump,3FWA*P2, discharge valves, 3FWA*HV36A, B, C, D at power levels less than 10 % was identified as l
a possible violation of Technical Specification (TS) 3.7.1.2. This TS states that, "At least three independent steam generator auxiliary feed-water pumps and associated flow paths shall be operable." Past plant operations have used the TDAFW and the Motor Driven Auxiliary Feedwater (MDAFW) pumps during startup, causing a portion of the TDAFW pump discharge piping to be subject to high pressure discharge from the MDAFW pumps during startup and shutdown evolutions. Being subject to the MDAFW pump discharge f
t pressure would require the TDAFW discharge piping to be classified as High Energy Line Break (HELB) piping, which it was not. The licensee elected to qualify the lines of interest as HELB lines. However, because of the lengthy process of qualifying the line as HELB, the licensee issued administrative controls to require that the Main Feedwater (MFW)
Pumps be used during normal plant startups and shutdowns until the lines of interest ,
were HELB qualified. The qualification of the TDAFW discharge piping as HELB lines, j implemented as a design change record (DCR), was evaluated during this inspectio I Observations and Findinas The inspector reviewed DCR M3-97046, Rev.0, along with associated documentation, I calculations and engineering drawings. The options originally considered by the licensee to correct the lack of HELB qualification of the discharge line were: 1) install the required barriers and make other changes required to qualify the line as an HELB fluid system line, 2) request a TS change allowing closure of the discharge valves 3FWA*HV36A, B, C and D when below 10% power,3) develop an engineering justification that the TDAFW pump
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discharge valves could be closed at power levels less than 10% without making a request to change the TS, and 4) use the MDFW supply pumps instead of the AFW pumps during plant startup and shutdown, leaving the TDAFW pump discharge valves ope The licensee revised procedures and completed training to allow the plant operators to use I the MDFW system during startups and shutdowns while taking steps to qualify the AFW lines as HELB as in Option 1 above. Design Change Record DCR M3-97046 contained the calculations, engineering evaluations and other materials required to qualify the AFW lines as HELB as well as procedure changes to allow valves 3FWA*36A-D to remain open during startu I DCR M3-97047 and other documentation established the HELB boundary as between l
check valve 3FWA*V31 and the four valves 3FWA* 36A through D. Later analysis by the '
licensee, based on USNRC NUREG-0800, Section 3.6.2, " Determination of Rupture
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Locations and Dynamic Effects Associated with the Postulated Rupture of Piping," Part !
l 1.3.a, entitled "At Terminal Ends," determined that the HELB terminal end should be the first normally closed valve. The first closed valve would be check valves 3FWA*V43, V39, V39, V35 and V47, which are upstream and close to their associated valves 3FWA*36A though D. Accordingly, the HELB sections of the TDAFW system, evaluated I
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by the licxsee, were the lines between the 3FWA*36A through D and the check valves 3FWA*V43, V39, V35 and V47. However, the documentation provided by the licensee stated in numerous places that the HELB boundary was between the 3FWA*36A-D and the check valve 3FWA*V31 and did not reflect the change to the other referenced check valves. The licensee agreed to modify the documentation to show that the HELB line no longer terminated at the 3FWA*V31. Moving the terminal end of the HELB line from 3FWA*V31 to the listed check valves, shortens the length of the HELB lines to be analyzed. The inspector reviewed the calculations and evaluations made by the licensee l showing that the lines in question were HELB qualified. No discrepancies were identified, l
beyond the agreed upon changes noted abov .C_qnclusions l
- Currently, the licensee has issued procedures to allow plant startups and shutdowns to be accomplished using either the MDAFW pumps or the TDAFW pumps. The discharge lines of the TDAFW pumps are now HELB qualified, allowing the piping and components to be subjected to the high pressure discharge from the MDAFW pumps. The actions taken by the licensee are deemed adequate in providing the plant operators the flexibility for
, conducting startup/ shutdown evolutions without violating the plant technical specifications
- or raising new TDAFW desigri concerns (reference SIL ltem 11, closed in inspection report ~
50-423/97-208).
- E2.3 Containment Foundation Erosion (Closed - SIL ltem 12) Inspection Scope (37550)
In inspection Report (IR) 50-423/94-11,tha NRC reviewed an issue regarding the erosion l j
of cement from the porous concrete drainage system installed under the containment
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basemat at Millstone Unit 3 (IFl 50-423/94-11-09). The licensee's efforts to follow up and resolve this concern were further documented in NRC IR 50-423/96-04. On October 18, l
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1996, the NRC staff requested additional information regarding the issue and the licensee's i plans to resolve the issue. This inspection documents the NRC's review of the licensee's submittals and evaluations since 1996. The NRC staff reviewed the licensee's submittals l to understand and assess the effects of the cement erosion process on the functionality of )
the containment for the near-term (to year 2000) and the long-term (the plant's licensed I life, i.e., year 2026). Inspector Followup Item (IFI) 94-11-09 was subsequently closed in IR 50-423/98-207,with the remaining technical aspects of this issue being tracked as part of Office of NRR review of SIL ltem 1 Observations and Findinas The containment structure at Millstone Unit 3 is a steellined reinforced concrete structure j consisting of a vertical cylinder wall with a hemispherical dorne and a flat circular basemat.
l The basemat is 10 feet thick, and is founded on essentially impervious rock. However, i between the rock surface and the underside of the basemat there are several layers. These l layers in ascending order are: (1) a 10-inch thick leveling layer of porous concrete made of j coarse aggregates and low alkali Type ll Portland cement; (2) a 1/16th inch butyl rubber j waterproofing membrane; (3) a 2-inch thick mortar seal; (4) a second layer of 9-inch thick i
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porous concrete made of coarse aggregates and high alumina cement (HAC); and (5) a thin mortar seal on the top of the upper layer of the porous concrete. Six-inch diameter porous concrete pipes are installed in the upper porous concrete layer to collect and drain ground water, which may seep down along the periphery of the containment wall. The collected water drains into two sumps inside the Engineered Safety Features (ESF) buildin In a letter, dated April 16,1998, the licensee provided a summary of various investigations performed to address the concerns raised regarding the cement erosion, initially (1995, 1996), the licensee tested mock-ups of the porous concrete media; subjecting them to extremes of water flows to simulate the wors' case scenarios (Phase 1, Phase 11, and Phase til programs). The primary objectives of these test programs were to determine the !
reasons for the observed white residue collected in the sumps, and any impact on the integrity of the subfoundation porous conuMe materials. Construction Technology I Laboratories (CTL-licensee's consultant) subseque:1tly performed a technical investigation; I (1) to determine the reasons for formation of the white deposits; (2) to establish a root cause mechanism for deterioration of the sub-base concrete, as reflected in the mock-up specimens; and (3) to provide an estimate of the residual strength of the in-situ porous concrete. These investigations indicated the primary concern to be the integrity of the high l alumina cement used in the upper porous concrete layer. This concern was not investigated by the mock-up test j l
Subsequently, the licensee decided to obtain actual HAC porous concrete cores from the i containment subfoundation to establish, conclusively, any degradation mechanisms and the in-place compressive strength of the HAC porous concrete material. The cores were extracted from the subfoundation of the ESF building. In a letter, dated December 19, 1997, the licensee provided the technical bases to justify the use of these cores as representing the HAC porous concrete under the containment. The cores were examined by CTL, and their strength evaluation was performed by Attran Corporation. The NRC staff l
reviewed these reports and other information provided by the licensee. The staff's review was broken down into four areas: (1) the effect of the quantity of the cement erosion from the porous concrete; (2) the effects of ' conversion' of the high alumina cement; (3) the adequacy of the existing porous media to transfer the containment loads to the bedrock; 4 and (4) the short-term and long-term functional integrity of the containment. The*te areas are discussed belo The Effect of Cement Erosion from Porous Media:
l in normal building construction on relatively pervious soils, the functions of the under basemat drainage, and the load transfer from the basemat to the foundation medium could I be accomplished by well compacted graded crushed stone (i.e., without any cement in the l
. crushed stone). However, on a relatively impervious foundation medium (as at Millstone j l Unit 2), the licensee's consultant decided to use porous concrete (and the embedded i I
pipes) to drain the accumulated water, and to transfer the loads to the bedrock, in this j case, the bearing strength of porous media must be sufficient to transfer the containment loads to the basemat without detrirnental settlemen In response to the staff's question on the loss of cement as a result of erosion from porous concrete, the licensee provided the following assessment in its December 19,1997, letter.
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l On the basis of the total dry weight of 752 puunds of residur. coilected from 1987 to 1996 I (9% years), the licensee estimated the dry weight of the cen ent residue collected from the cumps to be on the order of 80 pounds per year. ProratLT this data for 25 years from j 1975, the year of porous concrete placement, to the year 2000, the total quantity of eroded cement would be 2000 pounds. Considering the loss of cement at the rate of 100 {
pounds per year for 51 years (license's life for the unit, i.e., 2026), the total amount of
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cement lost would be 5100 pounds . The total cement weight in the porous media is j 670,000 pounds. This amount of cement eroded would be 0.3 percent by year 2000, and
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0.76 percent by the year 2026. Such quantities of cement loss would not by themselves adversely affect the functioning of the porous medi Based on the review of the licensee's reports, the cement erosion of the magnitude discussed above has not significantly affected the condition of the porous medi However, the water drainage should be effective, and the white residue in the sumps should be monitored, and its effects evaluated periodically to ensure that the amount of /
erosion is within the assumptions of insignificant " cement erosion." In a letter, dated April 27, !998, the licensee committed to (1) measuring the white residue / mass loss of calcium-aluminum in the ESF sumps, semi-annually, and (2) monitoring the containment structure settlement (external surface every 2 years, internal every 3 years). The second commitment w-ill be integrated with the licensee's condition monitoring program for safety +
related structures; in concert with the implementation of 10 CFR 50.65," Maintenance .
Rule." The program will also include the monitoring of heavily loaded areas for evidences of settlement. The NRC found these actions appropriate and had no further questions in i this are The Effects of ' Conversion' of the HAC* f in response to the staff's question on the possibility of conversion of the HAC, the licensee i attempted to incorporate the needed parameters in the ongoing Phase 111 mock-up testing in 1996. The conversion of HAC is of particular interest because it leads to a loss of strength due to the fact that the converted cubic C 3AH, hydrate has a higher density than the nonconverted hexagonal CAHw hydrate. Thus, if the overall volume of the paste structura f containing HAC is constant, conversion results in an increase in the porosity of the cement
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paste, which adversely influences the strength of HAC concrete. Based on the microscopic, macroscopic, and petrographical examination of the core samples from the Phase 1, Phase ll, and Phase lll mock-up tests, samples from the white residue in the sumps, rock samples from the underlying rock and glacial till (obtained from the bored wells), and analyses of the ground water and the sump water, CTL made a number of observations, but could not come to any conclusion regarding the extent of conversion of HAC or any other chemical effects on the porous media or the basemat concrete. CTL recommended that the utility should extract cores from the basemat and the porous media to understand the existing potential degradation of the porous media. The licensee's letter of December 19,1997, describes the locations of cores, the core product examination
- results, and the strength evaluation results. As documented in the licensee's letter of April l 16,1998, CTL concluded that (1) the high alumina cernent (HAC) porous concrete is in
' good condition and will continue to serve its planned function, (2) no significant chemical or physical degradation has taken place that negatively impacts the performance of the HAC concrete, (3) the potential is small for further changes to the chemical and physical i
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structure of the HAC paste, and (4) since the concrete is already 23 years old and is in a cool, stable environment, it should remain in good condition for several more decade Based on the review of the information provided by the licensee, the staff concluded that the HAC porous concrete media, and the ordinary Portland cement (OPC) media, have not undergone significant conversion or other chemica! reactions that would jeopardize the functioning of the porous media. However, the staff finds CTL's recommendation and the licensee's commitment that water (from the tvvo monitor wells instelled in the ESF building)
be sampled and tested periodically to track water chemistry changes and evaluate the potential effect on the performance of the HAC porous concrete appropriate. The NRC had no further questions in this are Load Transfer throuah the Existina Porous Media:
The design compressive strength of the two porous concrete layers is 1000 pounds per square inch (psi). The licensee had performed mock up tests to simulate the amount of erosion when the HAC and OPC concretes uere subjected to the accelerated water flows through them. The strength test performed on the cores from the Phase il mock-up test indicated that the average compressive strength of 69 samples of the eroded concrete was calculated to be 1258 psi, and the average of the lowest monthly test samples taken from the Phase lll mock-up testing was calculated as 877 psi. The later investigation of the l core-samples taken from the basemat of the ESF building, in confined condition, indicated j the existing strength of the HAC concrete as 2800 psi. Included in the licensee's l December 19,1997, submittal, a consultant to the licensee evaluated the strength test
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results against the containment loadings. In this assessment, the consultant considered the bearing pressures on the porous concrete media from the imposed containment loading to be 52 psi from the static dead and live loads, and 215 psi from the postulated load combinations incorporating the safe shutdown earthquake (SSE) load and design internal pressure load. A failure mode of potential concern was hypothesized, as follows:
Because of its porous nature, the concern is that the porous concrete might crush a significant fraction of its thickness if subjected to an SSE or accident pressure i loading thus leading to an unacceptable settlement of the containment buildin Potential problems could result if the porous concrete were to crush more than 2%
of its thickness. A very conservative acceptance criterion would be that the HAC l porous concrete were to crush more than 2% under a 650 psi plane strain loadin l This criteria [ sic] provides a factor of safety of 3.0 (= 650/215) against any potential problem which could result from crushing settlemen The consultant then used the results (stress-strain relationship) of the core tests to demonstrate that the imposed bearing stress on the HAC concrete medium would have to reach 5000 psi or greater before 2 percent shortening of this layer might be reached, giving a factor of safety of over 20 against the undesirable settlement of the containmen l The NRC staff agreed with this failure mode hypothesis and tne acceptance criterion, but i does not agree with the consultant's (1) strong assertions regarding the use of the I confined core test data, (2) disregard for the potential settlement of the 10-inch OPC j concrete medium, and (3) disregard for the possibility of some of the scenarios considered !
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l 69 by the licensee in its evaluations. However, on the basis of (1) the results of the worst case scenarios hypothesized by the licensee [as discussed in (4) below), (2) the
! consultant's assertion regarding a large margin against crushing of the HAC concrete rnedium, and (3) the fact that the containment has not undergone any settlement in the l
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!ast 23 years, the staff concurred with the licensee's conclusion that HAC and the OPC concrete media are capable of transferring the postulated containment loading to the bedrock without undesirable settlement. The NRC had no further questions in this are FunctionalIntearity of the Containment:
In its December 19,1997 letter, the licensee evaluated four hypothetical scenarios (1)
uplift under the SSE combined with the maximum hydrostatic head, (2) loss of integrity of porous concrete around the peripheral drain pipe, (3) loss of porous concrete integrity around the two interior drainage pipes, and (4) a loss of porous concrete integrity in five circular areas with a diameter of 5 feet each, in the vicinity of the drainage pipe intersection, together with the assumption that the drainage pipes are filled with the cement eroded from the porous concrete. In scenario (1), the licensee contends that a factor of safety of 1.5 exists against uplift of the containment. In each of the other three
. scenarios, the licensee estimated the amount of cement loss, the loss of the foundation -
bearing areas, and the ability of the 10 foot thick basemat to span the degraded areas of the porous concrete. Based on the evaluation of these scenarios, the licensee concluded that the containment structure was operable, fully qualified, and able to perform its safety functio In the above evaluation, the bearing pressure was calculated assuming a rigid mat and linear distribution of bearing pressure. EOE International (licensee's consultant) expanded the above evaluation to address the effects of a loss of cement within the porous concrete on the design basis of the containment mat. The EQE analysis provided in the licensee's letter of December 19,1997, specifically included the effect of (1) loss of porous concrete strength, (2) possible differential settlement, and (3) different porous concrete stiffnes The bounding cases of porous concrete strength and stiffness were considered. Evaluation of the containment mat was performed using the ANSYS Finite Element Computer Program. Since the purpose of the analysis was to evaluate the containment mat, the finite element mesh was developed to ensure accurate stress results within the mat. Other structures, such as the crane wall, orimary shield wall, and containment wall were modeled to achieve accurate load transfer from these structures to the mat. Two load combinations were considered; (1) Dead Load (D),1.5 times the postulated accident pressure load (P,),
and Accident Temperature (T,), and (2) D + P, +T, + SSE. Based on these analyses, EQE concluded that the design basis for the Millstone Unit 3 containment mat were satisfied for various scenarios for strength, stiffness, and settlement condition Moreover, based on the results of petrographic examination, X-ray diffraction analysis, differentia' ' canning calorimetry analysis, and the mercury intrusion porosimetry of core l samples (from the ESF building), CTL came to the following conclusions regarding the I condition of the basemat and various layers of the containment subfoundation (from the f licensee's letter dated April 16,1998).
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Basemat Portland Cement Concrete: This dense, OPC concrete is in very good conditio No unexpected chemical or physical degradation was observed based on microscopical examinations. Steel reinforcing bars were observed to be tightly bonded to non-carbonated Portland cement paste and to be free of corrosion. Given its current condition and exposure, this concrete should remain in very good condition indefinitel HAC Mortar (seal): This layer is in very good condition. This layer is well-bonded to both the OPC basemat concrete and the porous HAC concrete. There is no evidence of chemical or physical degradation within this layer at the interface with the overlying OPC concrete layer based on microscopic.1 examinatio HAC Porous Concrete: The upper portion of this layer is more dense and has fewer voids than the lower portion of this layer. HAC paste in the upper portion is generally darker and less fully converted than HAC paste in the lower portion. Based on the concrete mix design, thermal history and exposure conditions, and analyses by differential scanning calorimetry, microscopy, porosimetry, and X-ray diffraction, it is CTL's opinion that conversion and subsequent chemical reactions have been mitigated by continued hydration and cold temperatures so that there has not been an overali significant loss in strength in ,
the HAC porous concrete laye Portland Cement Mortar (seal): This layer, underlying the HAC porous concrete and resting on top of the butyl rubber membrane is in very good condition. There is no bond between j the Portland cement mortar layer and the overlying HAC porous concrete. Calcium I hydroxide has been leached from the upper two or three millimeters of the Portland cement morta Portland Cement Porous Concrete (10-inch layer resting on the bedrock): This layer, underlying the butyl rubber membrane,is not directly accessible for analysis and no cores j were removed from this layer for examination. A water sample was collected and '
analyzed. The pH and chemical composition indicate teaching of some calcium hydroxide from Portland cement paste in this layer. Such leaching is common in OPC concrete and is not deleteriou Based on the results of a number of analyses, evaluations, and detailed examinations performed to demonstrate the integrity of the containment basemat, as discussed above, the NRC staff concluded that Millstone 3 contaic;nent integrity is not significantly affected by the small amount of erosion of cement from the porous concrete subfoundation. The staff further concluded that the licenWs commitment to monitor the condition of the containment and the heavily loaded structures, through periodic inspections, for evidence of settlement is appropriate. The NRC had no further questions in this are Conclusions t
l Based on the review of the licensee's submittals, site audits, discussions with the i licensee's staff, and the NRC staff's independent assessment of the current situatinn, the j f staff concluded that the erosion of cement from the underlying porous drainage system has I not jeopardized the containment's ability to perform its safety function for the immediate I future. Moreover, through an in-depth evaluation of the present and future potential [
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degradation of the porous concrete media, the licensee demonstrated that the containment structure will maintain its ability to perform the intended functions throughout the licensed life of the plant (until year 2026), and beyond. Based on the review of the results of (1)
the investigations of physical and chemical properties of the porous media and their existing condition, (2) the detailed analysis of the containment basemat subjected to the j worst case postulated loadings, and (3) judgement of the knowledgeable experts, the staff concluded that the containment structure would be capable of performing its intended function. However, as a prudent measure, the licensee has established an additional monitoring program (in addition to the normal maintenance) as described in the licensee's submittal of April 27,1998. The staff determined that these actions are acceptabl Based on the above, SIL ltem 12 is considered close !
i U3 E3 Engineering Procedures and Documentation E CFR 50.59 Safetv Evaluation Process (SIL ltem 78 - Closed) Insoection Scoce (37001) )
In IR 50-423/96-201 an NRC Special inspection Team concluded that the licensee's safety ,
evaluations were prepared in accordance with 10 CFR 50.59 and adequately supported the l determinations that the subject changes did not involve unreviewed safety questions !
(USQs). The team found several examples or inadequate safety evaluations which were considered to be apparent violations of 10 CFR 50.59 (eel 50-423/201-02,04,05,06, 07,08). Significant item List (SIL) Item 78, "10 CFR 50.59 Process," was included on the Millstone Unit 3 significant items for restart list to review the licensee's corrective actions for these apparent violations, to review the licensee's corrective actions for one unresolved item (URI 50-423/93-07-07) involving diesel generator fuel storage capacity, and to evaluate the licensee's overall 10 CFR 50.59 safety evaluation process. The 10 CFR 50.59 safety evaluation process was reviewed by several NRC groups, including Region 1 J based inspectors, the headquarters project manger, the resident inspectors, NRC Independent Corrective Action Verification Program (ICAVP) inspection teams, and the ICAVP independent contractor, Sargent and Lund j t
This inspection included a review of the NRC inspections from January 1,1997, to the present (including the inspection reports generated by the NRC ICAVP team inspections)
and the ICAVP independent contractor (Sargent and Lundy) interim report, which was issued on May 8,1998, as they related to the licensee's 10 CFR 50.59 safety evaluation process. The inspection also reviewed the licensee's response, dated March 2,1998, to the NRC Notice of Violation and Proposed imposition of Civil Penalties, which was issued on December 10,199 The NRC notes that on March 1,1998, the licensee superseded Nuclear Group Procedure (NGP) 3.12, " Safety Evaluations," with Regulatory Affairs and Compliance Procedure (RAC) 12, " Safety Evaluation Screens and Safety Evaluations." Both procedures have been reviewed during NRC inspections and the following summary documents the staff's overall review in this area.
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72 Observations and Findinas The licensee's corrective actions for the six apparent violations identified in IR 50-423/96-l
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201 were reviewed in three inspection reports. In IR 50-423/97-203,the NRC reviewed eel 50-423/96-201-02and 07, in IR 50-423/97-208the NRC reviewed eel 50-423/96- )
201-04, and in IR 50-423/98-206the NRC reviewed eel 50-423/96-201-05,06, and 0 I As discussed in these inspection reports, the NRC concluded that the licensee's corrective actions azequately addressed the technical aspects of the issues and the individualitems were considered closed. However, the NRC noted that the Notice of Violation (NOV -
l letter unique identifier 01232,01152,03082,01242,05014, and 01222) for each issue would remain administratively open. As part of these inspections, the NRC also reviewed Revisions 9 and 10 of NGP 3.12 and the associated training to implement the changes and stated that the procedure adequately defines the process for preparing safety eva!uations for proposed plant changes to assure safety and coenpliance with 10 CFR 50.59 requirements. The NRC stated that the procedure is clear and includes screening check sheets to simplity and standardize the process. The NRC further concluded that lesson plans for the associated training fully described the 10 CFR 50.59 requirements, their intent, and the methods of meeting them. The NRC noted that an important addition to the l enhanced 10 CFR 50.59 process appeared to be the requirement to have the Plant and/or-Site Operations Review Committees (PORC/SORC) review all safety evaluations; this additional review by knowledgeable personnel adds a significant quality check to the process.
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The licensee's corrective actions for URI 50-423/93-07-07were reviewed in IR 50-423/97-207. The NRC concluded that the revised procedures contain adequate guidance to ensure that the licensee will have adequate and reliable fuel oilinventory in the storage tanks for seven days of continuous diesel generator operation following a loss of power or loss of coolant accident at Millstone Unit 3.
l As documented in IR 50-423/98-206,the NRC randomly selected and reviewed 10 additional 10 CFR 50.59 safety evaluations for a more detailed review. The NRC reviewed the plant changes to ensure that the proper safety evaluation was completed, that the ( licensee adequately addressed the three questions in 10 CFR 50.59, and that the licensee followed its own guidance. For the safety evaluations reviewed, the inspector determined that the licensee ac0quately addressed 10 CFR 50.59 and, therefore, no USQs were identifie The NRC ICAVP Tier 2/ Tier 3 inspection also addressed the 10 CFR 50.59 safety evaluation process in IR 50-423/97-209. The ICAVP team reviewed NGP 3.12 and found a number of inconsistencies between the procedure and 10 CFR 50.59. Since the problems with the threshold for writing a safety evaluation had been previously identified by the licensee, the team also reviewed the licensee's draft procedure to replace NGP 3.12, which was procedure RAC 12. During the inspection, the licensee committed to revise RAC 12 in a number of areas based on findings of the team. During a followup inspection, the NRC reviewed the changes made to RAC 12 and found them adequat Therefore, the NRC concluded that the licensee's 10 CFR 50.59 change process was acceptable and met the requirements of 10 CFR 50.59.
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During the implementation of the ICAVP, Sargent and Lundy reviewed the licensee's processes for making changes to the Millstone Unit 3 licensing and design basis, in the l interim report on the ICAVP, Sargent and Lundy stated that the design of plant I modifications impbmented after receipt of the plant's operating license was technically l adequate and configuration control was adequately maintained. Sargent and Lundy further stated that the licensee has established programs, processes, and procedures to maintain effective configuration control of their licensing and design basis in the futur The NRC reviewed the licensee's Reply to a Notice of Violation and Proposed Imposition of l Civil Penalties, dated March 2,1998, as it related to the 10 CFR 50.59 safety evaluation process. The corrective actions listed have been reviewed during the inspections
! previously mentioned. As such, the NRC staff finds the licensee's corrective actions i acceptable.
I Conclusions Based on reviews and inspections noted above, the inspector concluded that the licensee's 10 CFR 50.59 safety evaluation process is adequate and should maintain effective configuration control of the Millstone Unit 3 licensing and design basis. Although some issues with the implementing procedure were identified, the inspector determined that the ]
licensee's corrective actions and revised procedure RAC 12, adequately control changes to i the facility. Therefore, as a result of this inspection, along with the results of the other inspections of the safety evaluation process documented above, SIL ltem 78 is hereby closed. The NRC Notice of Violation (NOV -letter unique identifier 01232,01152,03082, 01242,05014, and 01222) included in the NRC letter dated December 10,1997, can also be close U3 E7 Quality Assurance in Engineering Activities E Response Time Testino (Cbsed - SIL ltem 9)
(Closed) LER 50-423/98-06: Failure to Perform Response Time Testina for Relavs Inspection Scope (92903)
The inspectors reviewed licensee actions taken in response to ACR 10543 which identified the need for additiona! evaluation of procedures to ensure all components are included in the response time surveillance test Observations and Findinas )
To determine the response time of reactor trip and engineered safety features actuation system (ESFAS) functions, component response time data are collected by the instrumentation and controls, maintenance and operations departments. The total time for a particular function is then obtained by summing the sensor, instrument rack, actuation logic and final component resoonse times. This summation is performed in procedure SP j 31024," Calculation of Reactor Trip and ESF Response Times," and the total time for the
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function is compared to the acceptance criteria in the procedur _ _
in March 1996, the licensee identified that the response time of the slave relay in the turbine driven auxiliary feedwater pump automatic start circuit had not been considered when evaluating the acceptability of the overall circuit time response. This issue was documented in ACR 10803. ACR 10543 was subsequently issued to perform a broader evaluation of the response time testing procedure As a result of this evaluation, changes were made to the turbine driven auxiliary feedwater pump surveillance test and the slave relay surveillance tests to ensure response time data are collected for all of the slave relays. Procedure SP 31024 was also revised to ensure slave relay data were included for all component During an initial NRC review of this issue in 1997, the licensee indicated that additional improvements to procedure SP 31024 were appropriate to avoid future problems with the collection and evaluation of response time test data. The licensee then issued CR M3-97-3089 to initiate further investigation and to review the deportability of the condition. The licensee concluded that all of the relays were included in at least one component time response and that the test method was sufficient to ensure a failed relay would be identified during testing, in January 1998, during an independent review of procedure SP 31024, the licensee identified two additional deficiencies in response time testing. The deficiencies included relays in the circuit path that results in an EDG start on a safety injection (SI) signal and sensor relays and auxiliary relays associated with the degraded grid voltage start signal for the EDG. These deficiencies were reported to the NRC in Licensee Event Report (LER) 98-006-00 and the corrective actions included the following:
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- The affected surveillance procedures were revised and performed to include the affected components for degraded grid circuitry and the EDG start on an SI signa * Ownership of the response time testing program was established within the technical support departmen The inspector reviewed the licensee's corrective actions and found them to be appropriat This licensee-identified technical specification non-compliance is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Poliev. LER 50- j 423/98-06-00is considered close l The inspector reviewed Revision 3 of SP 31024 and found that the procedure included a l response time test of the power range neutron flux reactor trip high set point channel but
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the power range neutron flux reactor trip low set point channel was not specifically included. The licensee found that both the high and low trip circuitry was being time response tested but the summation test required clarification to use the longest time when i performing the summation. Also, containment isolation valve 3DAS*CTV24 was included in previous revisions of the test but was inadvertently omitted in Ravision 3 of the procedure. The licensee documented these discrepancies in CR M3-98-137 I On March 14,1998, the licensee approved Revision 4 to procedure SP 31024 which )
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corrected the deficiencies noted above (CR M3-98-1374). The inspector reviewed the
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changes to the procedure and the results of response time testing performed usin9 this revision. No additional problems were identified, Conclusions The inspector concluded that the licensee adequately addressed time response testing procedure issues. SIL ltem 09 is close U3 E8 Miscellaneous Engineering issues E (Closed)IFl 98-82-03: Include calculation in alternate test olan to comoare unwedaina data to EPRI PPM hand calculation method
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The NRC Safety Evaluation regarding the EPRI PPM includes a condition that PPM users compare unwedging data to the PPM hand-calculation method for predicting unwedging thrust requirements. In a letter dated April 25,1998, NU committed to incorporate the comparison data into the Calculation 89-094-01513," Dynamic Test Results", and to j revise the ATP and dynamic test evaluation (DTE) forms to reference.the calculation. The l
inspector verified that the licensee commitment (B17178-08)had been completed by re/ ewing the revised Calculation 89-094-01513, ATP and DTE forms. This item is close l
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E8.2 [ Closed) IFl 98-82-04: Capture dynamic test data for cate and alobe valve load j sensitive behavior NU load sensitive behavior assumptions include a bias margin of 5.6%, and a random margin of 26.4% which is combined with other random errors using the square-root sum of the squares methodology. NU based these new assumptions on results published by EPRI as part of the Performance Prediction Program (PPP). NU also performed a statistical ana!ysis of in-plant testing that supported the use of EPRI's load sensitive behavior value Howevar, the inepectors noted that the licensee's use of the PPM in lieu of dynamic testing resulted in a limited amount of data available to support other program justification Therefore, the licensee will be expected to augment its load sensitive behavior analysis as i in-plant dynamic tests are performed in the future, in a letter dated April 25,1998, NU l committed to include the plant-specific load sensitive behavior information in Calculation 1 89-094-01513," Dynamic Test Results", and to reference the calculation in the ATP and the DTE forms. The inspector verified that the licensee commitment (B17178-09)was completed by reviewing the revised Calculation 89-094-01513, ATP, and DTE forms. This item is close l l
E8.3 (Closed) IFl 98-82-05:Incornora_te in-olant stem friction coefficient data into an enaineerina calculati_oa l NU applied a worst-case stem friction coefficient assumption of 0.20. NU based this l assumption on reviews of Millstone-specific closing test data. The licensee statistically l
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analyzed using a 95% confidence level of in-plant stem friction coefficient data which resulted in a value slightly above 0.18. The inspector noted that the analysis was based on a small amount of test data. Therefore, the licensee will be expected to augment this ,
data analysis as part of its long term MOV program to increase confidence in the program's !
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stem friction coefficient assumptions, in a letter dated April 25,1998, NU committed to include stem friction coefficient data analysis in Calculation 89-094-01513and to reference the calculation in the ATP and DTE forms. The inspector verified that the ,
licensee commitment (B17178-10)was completed by reviewing revised Calculation 89-094-01513, ATP, and DTE forrns. This item is close {
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E8.4 (Closed 1JFl 98-82-08:Verifv aoorooriate resolution of the 110% toraue limit issue !
betwcen Limitoroue and NU During review of NU's thrust calculations, the inspector noted cases where toraue at torque switch trip was allowed up to 110% of the actuator's torque rating. In a letter dated April 25,199'8, NU stated that it had contacted Limitorque to ensure the acceptability of its approach. During this inspcction, the inspector noted that Limitorque has indicated that licensees should ensure that the applicable structural limit (which may be greater than rating based on specific studies) is not exceeded when considering torque j switch repeatability and diagnostic accuracies. NU agreed with this clarification. This item l is close I l
E8.5 (Closed) IFl 98-82 09- Verify a_Qceotable resolution of Yarway alobe valve issues p_rior to restart Two Yarway globe valve problems existed during the time of IR 50-423/98-82. The first problem involved the RHR Pump Miniflow Recirculation valves (3RHS*FCV610/611). The i
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licensee found that these valves wiii repeatedly cycle open and close during surveillance testing due to the design of the flow control circuit that governs operation of the MOV Therefore, a modification was initiated to modify 3RHS*FCV610/611'scontrol circuit The inspector verified that the modification was cornpleted and CR M3-97-4475,which described the problem, had been closed. The second problem, which was described in CR M3-97-4541, involved a licensee-identified concern regarding a potential for breach of the va!ve pressure boundary in the drive open condition. The inspector reviewed the licensee's evaluation and testing performed to justify the adequacy of the MOVs susceptible to this potential problem. The inspector agreed that the licensee's basis for i closing CR M3-97-4541 was reasonable. This item is close E8.6 (Closed) VIO 98-82-10: Correct MOV thrust calculational errors Individual MOV thrust calculations are performed in accordance with Calculation #97-MOV-01012MG, " Technical Justification / Methodology for Preparation of Millstone Units 1,2, &
3 MOV Thrust / Torque Calculations," which provides a QA method for performing such calculations in accordance with PI-9, " Determination of Stem Thrust Requirements." NU used a database software tool (i.e., Smartbook) to store data and to develop the MOV target thrust / torque calculations that were used to establish design-basis requirement During IR 50-423/96-82 the inspector identified several errors in specific MOV calculations such as Calculation 89-094-1017ES," Millstone Unit 3 Target Thrust / Torque Calculations for 3SIL*MV8808A,3SIL*MV8808B,3SIL*MV8808C,3SIL*MV8808D," Rev. 4, dated January 26,1998, and Ca!culation 89-094-0900ES," Millstone Unit 3 Target
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Thrust / Torque Calculation for 3SIL*MV8804A,3SIL*MV8804B,"Rev. 5, dated February 9,1998. The errors included: The minimum required opening thrust was incorrectly compared to the actuator's open structural limit instead of the actuator's open degraded voltage thrust capability (which was more limiting); The estimated dynamic disc pullout thrust requirement (based on the EPRI PPM unwedging calculation) was incorrectly compared to the actuator's open structural limit instead of the actuator's open degraded voltage thrust capability (which was more limiting); An incorrect torque value was sometimes displayed in the comparison answer box; the software did not accurately select the minirnum torque limit as required; and In some cases, the software truncated the most significant digit of the torque value displayed in the comparison answer bo Given the significance and generic nature of the errors found in design-basis calculations that had received an independent second-level engineering review, the NRC considered these errors to represent several examples of a potential viciation of 1CCFR50, Appendix B, Criterion lil, " Design Control," which requires that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions, in its letter datad April 26,1998, NU agreed with the NRC findings. The licensee stated that MOV project personnel had not validated Smartbook, that past changes to PI-9, the methodology calculation, and Smartbook had not been adequately controlled, and that MOV project engineers had demonstrated inadequate attention to detail in perferming the calculations. The licensee also made the following commitments: B17178-01: A comparison and consistency study was to be completed by May 5, 1998, among PI-9, the methodology calculation, and Smartboo . B17178-02:MOV engineers using Smartbook were to be retraine . B17178-03: All Unit 3 MOV thrust calculations were to be reviewed and revised by )
May 5,1998, to conform to MOV program design methodolog l B17178-04: A new MOV Department Instruction was to be developed by June 30, !
1998, to administratively control changes needed to continuously u.odate l Smartbook and ensure consistency with methodology calculation and MOV program l PI- j
) B17178-05: All MOV calculations and MOV program documents were to be revised prior to May 5,1998, and restar l l
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t I During this inspection, the inspector verified that NU had completed the CR M3-98-0790 Level 1 root cause analysis. The licensee had finalized the 2 basic root causes as being:
(1) Organization changes caused a lack of consistency among Pl-9, the methodology calculation, and Smartbook; and (2) Complacency by engineers in the preparation and independent review of thrust calculations. Based on a sample review of the revised thrust calculations, the review of MOV personnel training records, and discussions with personnel, the inspectors determined that the licensee had adequately completed the commitments described in the letter of April 25,1998. Violation VIO 50-423/98-82-101s close E8.7 (Closed) URI 98-82-11: MOV Pronram Qualification of Personnel- Trainina Requirements During IR 98-82, the licensee determined that the records for some specific training of MOV personnel had been lost. This item had been opened to perform additional followup inspection to assure that this documentation problem was an isolated occurrenc !
In a letter dated April 25,1998, responding to this item, NU provided a brief discussion indicating that a thorough review of all training records produced no other deficiencies. NU concluded that the prior missing records was an isolated occurrence. The inspector discussed the essence of this detailed review and found that NU's conclusion was appropriate. However, the inspector learned that an additional administrative detail problem regarding MOV personnel training records had been found and had not been included in the letter of April 25,1998. Since this new information was pertinent to MOV training documentation problems, the inspector requested NU to submit an updated response regarding this ite As documented in a supplemental response letter of May 22,1998, regarding the additional administrative detail problem found, NU learned that the current MOV program training qualification requirements did not include Position Specific Training qualification records to identify the various MOV program instructions applicable to the tasks that were performed. As a result, not all MOV program instructions pertaining to MOV program tasks were identified in training records. Qualification of MOV personnel was not a concern since prior interviews and discussions evidenced to the inspector that the MOV personnel in place were qualified to perform their job responsibilities. To establish the long term administrative control over the training that is required in the MOV program and to prevent recurrence of missing training files, NU committed (Commitment No. B17260-01)
to the following action to be completed by June 30,1998:
The MOV Program will develop Specific Position training to be captured in a Training Qualification Record (TCR) per the requirements in NTM 7.202,
" Engineering Support Training implementing Procedure." The programmatic controls involving the use of the TOR and the Specific Position Training requirements that
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apply will be incorporated into the MOV Program Manual.
l The aforementioned training records problem is a f ailure that constitutes a violation of minor significance and is not subject to formal enforcement action. Unresolved item URI 50-423/98-8211is closed.
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E8.8 LER 97-46-00 Containment Recirculation Sorav System Cubicle Flood Potential frpoection Scone (37550)
The inspector reviewed the actions taken by the licensee to resolve the concern with groundwater leakage into the recirculation spray system (RSS) pump room sump Observations and Findinas in May 1997, a licensee configuration management program review identified that groundwater inleakage to the RSS pump cubicles in the Engineered Safety Features (ESF)
building could impact the operability of the RSS pumps and the primary containmen The original plant design included a rubber membrane that encased the containment substructure and was intended to prevent any groundwater effects on the containmen (For a more detailed discussion of this issue, see paragraph E.2.3 of this report.)
Groundwater that leaks past the membrane is directed to the RSS room sumps by perforated pipes located in a porous concrete layer under the containment basemat. The original design assumed that little or no groundwater would bypass the membran However, the licensee identified that there is approximately 1,000 gallons per day of groundwater leakage into the sumps due to an apparent failure of the membrane. During normal operation, the groundwater is removed from the sumps by non-safety sump pump The f ailure to remove the groundwater would result in unanalyzed hydrostatic effects on the containment liner and eventual flooding and failure of the RSS pump motors. The RSS pumps are required for long term core cooling following an accident. Since the RSS rooms are not accessible post-accident due to postulated high radiation levels, and the installed pumps are not safety-related, the licensee decided to install a safety-related sump pump system. The new system, except for sharing the same sump, would be totally independent of the original system and would only be used during post-accident condition j l
The basic design and operation of the system included the installation of safety-related I pumps, valves and piping from the sumps to the outside of the ESF building. The pumps I are centrifugal pumps driven by an air motor. Air to power the pumps is provided by non- 1 safety portable air compressors located outside of the ESF building and connected to the !
safety-related portion of the system by hoses. The discharge of the sump pump is directed i to collection tanks through a hose that connects to the safety-related portion of the system. One aspect of the design bases of the system is that it must remain operable for a year following an acciden The sump pumps were purchased as commercial grade components and then upgraded, by l
testing and analyses, to safety-related components. The upgrade was performed by a j vendor who then provided the qualified pumps to NNEC The inspector reviewed the plant modification to determine if the system was operable to perform its design function. The licensee had determined that the groundwater leakage condition constituted an unreviewed safety question and submitted a license amendment that will be reviewed by NR _ _ _ _ _ _ ____
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The inspector's review included a review of the plant modification and related documentation including:
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- design change record (DCR M3-97079),
- qualification test report,
- post-modification test procedure and results,
- pump vendor manual,
- system operating procedures, and;
- planned surveillance testing and preventive maintenanc The inspector also performed a walkdown of accessible portions of the system with the design enginee The inspector identified the following conditions that would potentially affect system operability: There was no air filter installed between the air hose connection and the safety- l related portion of the system. This would allow any dirt or debris from the non-safety air source to collect and eventually clog a filter located at the pump air motor inlet which cannot be accessed for cleaning. The inspector noted that the vendor manual recommended the installation of an air line filter even for commercial application . There was no safety-related overpressure protection for the pump air motor. The I pump nameplate states the maximum operating pressure is 90 psi. There was a I relief valve installed on the air compressor unit but it was not safety-related and the I setpoint was at 150 ps . The pump qualification report addressed the post-accident radiation effects on three o rings in the pump but did not evaluate the shaf t seal or a rubber coated gaske . The qualification report did not evaluate potential corrosive effects on the air motor and did not provide any necessary operating guidance when taking the pump out of service. The vendor manual contained specific directions to drain and oil pump if taken out of service, l The test report documented pump failures during mechanical qualification tests. it l was not clear that there was adequate qualification testing to ensure long term performance of the pump . The vendor manual states that a lack of lubrication will result in rapid pump failur Although proper air oiler operation is important to pump operation, the surveillance )
l and operation procedures did not verify oil delivery rat l l Due to the long vertical pipe runs the inspector questioned whether there wouN be a concern for gradual oil accumulation in the ai. piping due to oil addition during preventive maintenance and during operation of a:r oiler, j i
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! 81 The air motor contains a governor function that was not specifically tested during qualification or post-mod testing, and it was not clear if the governor was necessary to prevent pump damage due to over spee The licensee documented these concerns in CR M3-98-2932 and subsequently took the following immediate actions to resolve questions associated with the sump pump design adequacy:
The system was modified to install an air strainer in the safety-related portion of the f
system at a location that will be accessible following an acciden j
The design air pressure for the system was reduced and a safety-related relief valve !
was installe * Additional evaluation of the non-metallic parts was performed to determine their ability to withstand the postulated radiation effects that would exist following an accident. The gasket questioned by the inspector was determined to not be acceptable and was replaced with a suitable material. The pump shaft seals were found to be acceptabl * To prevent corrosion of the air motor, the licensee plans to implement a preventive maintenance task which will periodically add oil to the air inlet to the pump and run l the pump at an idle speed to ensu, all surfaces are coated with oil. This will be performed with the air outlet pipe disconnected to prevent excess oil from j accumulating in the piping. Further review by the licensee revealed that the lubricating oil that is specified for use in the system includes a rust inhib tor and the vendor specification indicates that the oil provides excellent rust and corrosion protectio ]
- The licensee had the vendor that performed the qualification testing perform j additional tests. This included testing that subjected the pump greater than the number of start and stop cycles that would be required for the pump to parform its function for one year following an accident. During this testing the licensee experienced problems with the air motors due to excessive wear of the metal rotor blades. The metal blades were replaced with non-metallic (Hylum) blades and satisfactorily tested by the vendor. The installed pumps were subsequently reworked to replace the rotor blades with inc same type that performed satisfactorily during the qualification testin * The in-line oiler in the air supply was upgraded to a safety-related component and oiler operation was clarified in the procedure * The licensee determined that the govemor function of the air motor was necessary to ensure proper operation. The cyclic testing discussed above demonstrated long l term operability of the governo In addition to these actions the licensee formed an event review team to review these l issues and to assess whether other modifications may have similar problems. The licensee j
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concluded that the issues identified with this modification were an isolated problem due to the nature of the modificatio Conclusions The inspectors concluded that the initial sump pump system design and the pump qualification were inadequate. Significant corrective action, including system and pump design changes, were required to ensure the system would perform its design functio The inspectors also concluded that the immediate corrective actions taken by the licensee l were adequate to ensure system operability. This item is unresolved pending NRC review
!. of the final results of the event review team, NRC review of any LER revision to be
! submitted and NRR review of the license amendment. (URI 50-423/98-208-07)
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L IV Plant Sunoort ( (Common to Unit 1, Unit 2, and Unit 3)
R1 Radiological Protection and Chemistry Controls Insoection Scope (83729)
.The inspector reviewed the licensce's programs for: (1) inventory and leak testing (where applicable) of radiological sources; (2) records required for the safe decommissioning of facilities under 10 CFR 50.75(g); and, (3) review of certain radiological wor In the area of source inventory and leak testing, a review of the licensee's records and data bases of radioactive sources was performed. A random sample population of sources was selected to verify storage location, physical description, marking and labeling of sources and source storage locations, as required under 10 CFR 20.1801. Also verified were the licensee's records for conducting leak testing of certain sources, as specified in the Unit 2 Technical Specification (TS) 3/4.7.7 and the Unit 3 TS 3/4.7.1 .
In the area of records required under 10 CFR 50.75(g), a review of records of spills or
' other unusual occurrences involving the spread of contamination in and around the facility, equipment, or site where significant contamination remains after cleanup was conducte In the area of radiological work, direct inspection of work being performed at the site during the period of May 18-22,1998 was performed. The review included survey data and ALARA documentation for work performed in the Radiologically Controlled Areas (RCA) and discussions with the Unit ALARA Coordinators on work planning activities for remaining work. This aspect of inspection also included a review of work controls, including briefings, radiation work permit, health physics control of work activities and maintaining occupational exposures for this work as low as is reasonably achievable (ALARA).' Particular inspection focus was made on high and locked high radiation area
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controls.
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83 Observations and Findinas Source Accountability and Leak Testina The licensee's program for the control of radioactive sources is defined by Procedure RPM 5.3.4, Revision 3, " Sealed Source Inventory and Control Program." A current inventory of all non-exempt sealed sources was maintained, including the results of semi-annualleak testing in accordance with Technical Specifications. The results of the most recent leak tests, conducted in December 1997, indicated that all sources were accounted for and that source integrity was maintained. The records confirmed that the licensee immediately removed a sealed source from service when leak test results indicated contamination in excess of regulatory limit The results of the licensee's most recent audit of sealed source control, conducted January 1998, identified three administrative discrepancies. Subsequently, the discrepancies were developed into Condition Reports to document the matters, effect appropriate corrective action, and track resolution. No recurrence of these discrepancies or similar matters was noted during this inspectio Decommissioning Records in Accordance with 10 CFR 50.75(al The licensee has records tn track spills or other potential releases of radioactive materials both within the owner controlled area and within the protected area. Within the owner controlled area, extensive surveys and analyses have been performed during the past 18 months analyzing areas where potentially contaminated soils could have been placed around the site and its environs. Detailed records of radiological surveys and environmental samples taken from various athletic and recreational facilities were availabl Results indicated that no licensed materials were present at these locations. Additionally, the State of Connecticut has also taken numerous measurements and samples of these facilities, independent of the licensee. No results from the state's analyses were available, ,
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Records of spills and releases of radioactive materials within the protected area were under development. The licensee has conducted a database search of plant records to identify ,
pot ' Mial events which could have resulted in contamination of facilities, and has compiled a list 1g of the results which were reviewed. Full documentation of these events, as j described in 10 CFR 50.75(g), was still under developmen Conduct of RadiW*a1 Work During the period at .
8-22,1998, the most significant radiological work was being conducted at Unit 3, and involved the repair of a leaking valve in containment (3RCS*V132). Appropriate work planning was being implemented to maintain occupational exposures ALARA for this project. A review of job controls in-place during I industrial radiography work in the Unit 3 Containment on May 19,1998 included postings and access control to the containment during this work.
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Allinspected areas at the three units and supporting facilities were determined to be appropriately posted and controlled in accordance with 10 CFR 20 and plant technical specification Conclusions The licensee established and implemented effective radiological protection programs with respect to: (1) control and leak testing of sealed sources and maintaining an inventory of these sources; and, (2) maintaining decommissioning records required under 10 CFR 50.75(g) for areas located outside the protected area. Decommissioning records for areas within the protected area are still under developmen Radiological controls were determined to be appropriate, especially in the areas of posting and control of high and locked high radiation areas, at all three units. Appropriate work planning for maintaining occupational exposures ALARA was also observed at Unit 3 for work on the 3RCS*V132 valve and for industrial radiography taking place in the containmen R5 Staff Training and Qualification in Radiological Protection and Chemistry Insoection Scope (83729)
Review of the licensee's continuing training program included selected training documents, lesson plans and training objectives, interviews with cognizant health physics and training personnel and direct observation of one session of mock-up trainin Observations and Findinas The licensee implemented a mock-up training program utilizing a small area in the Unit 1 Condenser Bay in late 1997. This facility takes advantage of the presence of some abandoned in place equipment and the current low (essentially background) radiation dose i rates in this area. Through the use of this equipment and personnel from the unit Health 1 Physics Department demonstrating the improper performance of various tasks, the licenseo has trained a number of supervisors and other site workers on the identification and immediate corrective actions to be taken for a variety of situations involving industrial, plant and radiological worker safety. One session of this training was observe Continuing training for health physics technicians is provided several times each calendar year. For 1998, the licensee recently completed its spring training cycle which included five days of training on statistics, spectroscopy, instrument operation, industry events and other associated topic Conclusions The licensee has established an effective continuing training program for plant workers and l radiation protection technicians, including the use of a detailed mock-up facility. Lesson l plans and training objectives leviewed were appropriate with regards to subject scope and depth of presentation.
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R7 Quality Assurance in Radiological Protection and Chemistry Activities Inspection Scooe (83722)
Licensee records of audits and appraisals in support of the annubi requirement to review the radiation protection program content and implementation in accordance with 10 CFR 20.1101(c) and self-assessments performed by various members of the radiation protection program during the past twelve months were reviewed. Audits, appraisals and self-assessments performed in the health physics area in 1997 and 1998 were reviewe (
1 Observations and Findinas !
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The licensee's program for meeting the annual audit requirement contained in 10 CFR i 20.1101(c)has been met in previous years through a combination of audits and appraisals I conducted by Quality Assurance Services (OAS), the Radiological Assessment Branch (RAB), and to a lesser extent each unit's radiation protection department. During the past six months, the licensee has developed a new audit and appraisal matrix for the years 1 1998-2000which includes work to be performed by the audits group of Nuclear Oversight l (the successor to OAS), and the Plant Support and individual unit Health Physics i Department Each audit reviewed was performed by qualified lead auditors or appropriate technical specialists. All findings and deficiencies identified were entered into the licensee's l condition report tracking system. Self-assessments are performed by all three units and I the plant health physics support organization of various aspects of their radiation l protection programs. Each unit's management has committed to the performance of a l
number of these assessments in addition to supporting the annual audit process required j under 10 CFR 20. While the scope of these assessments was smaller than the site-wide 1 audits, the technical depth of these reviews was appropriate. Deficiencies and findings from these assessments were verified as being incorporated into the condition report i program for documentation, tracking, and resolution of the issue Conclusions The licensee is implementing an effective program for conducting annual audits and appraisals to meet the requirements of 10 CFR 20.1101(c). Additional self-assessments are also performed by the various health physics organizations at the sit P8 Miscellaneous Emergency Preparedness issues P Unit 3 PASS Inspection (Closed - Unit 3 Sil. Item 83) Insoection Scoce (92904)
An NRC inspection was conducted on February 23-26,1998 in which a Severity Level lli Violation was cited regarding failure to adequately maintain the Post Accident Sampling System (PASS). (NRC inspection Report 50-423/98-01). During the periods of May 18-21,1998 and June 19-22,1998, an inspection was conducted to determine the adequacy
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of the licensee's corrective actions. These actions were to be completed prior to entering Mode 2. The inspection included a review of PASS procedures, surveillance test records, analytical results and interviews with engineering and chemistry personnel. Also, the inspector observed an EP PASS Drill and two PASS surveillance tests, Observations and Findinas While conducting the inspection during May 18-21,1998, the inspector determined that although the licensee had made improvements in their PASS program, they were not able to fully demonstrate the system was operational. Revised procedures had not been formally approved, system operability problems were still being investigated and an EP PASS drill was discatinued due to a mechanical failure. Therefore, the inspection was discontinued until crie 19,199 Subsequently, extensive system testing was performed to establish confidence in equipment performance and its reliability. Equipment deficiencies encountered have been correcte The inspector observed the collection of a containment air sample and a reactor coolant -
sample using the PASS. The samples were successfully collected and analyzed, and the liquid sample results were compared to a daily reactor coolant grab sample. The licensee intercompared the data to determine pass / fail based on commitments made in the Updated Final Safety Analysis Report (UFSAR). The inspector reviewed the licensee's analytical results from March through June 18,1998. All six sample points were tested at least twice. The 12 containment air surveillance tests were performed successfully and met the acceptance criteria. The 19 reactor coolant liquid samples were taken and all boron and chloride results were satisfactory. The sample results that were in disagreement prior to June 17,1998 were either the radioactivity, total dissolved gas and/or pH. The apparent causes for the disagreements were found to be dilution errors, sampling techniques or inadequate laboratory equipment. After correction of those issues, from June 16-20,1998 (six samples) all results were in agreement with the exception of the total dissolved gase The licensee is continuing to assess tha problem with determining the total dissolved ga Test results indicated they vvere able to consistently measure the dissolved gases within 30 cc(STP)/kg of the known value. However, they are committed to 10 cc(STP)/kg in their UFSAR. The licensee requested a licensing basis change regarding this matter in a letter (B17240) dated June 25,1998 to the NRC. Also, they have submitted an action plan describing their ongoing investigation into this matte Revised / approved PASS procedures were found to be detailed and thorough and provided the necessary steps needed for the chemistry technicians to prcperly operate the DAS The licensee has provided extensive training on the PASS to eight chemistry technicians, six of which were observed by the inspector. They were found to be knowledgeable of the l surveillance procedures and PASS configuration The licensee conducted an EP PASS drill in accordance with their Emergency Plan (E-Plan).
The drill adequately demonstrated that a PASS reactor coolart sample could be
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successfully collected and analyzed. Analytical sample results were provided to the Manager of Radiological Dose Assessment within the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> time requiremen One of the contributing factors to the PASS violation was a lack of management attention and oversight. Management has assigned the overall ownership of the PASS program to a Technical Support Engineering team and the ownership of the operation of the PASS has been assigned to the Chemistry Department. The Millstone E-Plan has been revised to include one PASS drill per year per unit. Also, the licensee has added to their Maintenance
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Rule Plan semiannual testing of a sump liquid and quarterly surveillance tests on reactor coolant and containment air for continuing system maintenanc Conclusions The inspector concluded that the corrective actions are sufficient to provide reasonable assurance that the PASS system would be able to assist in the assessment of core damage, given a significant transient or accident. Significant improvements were made in the PASS program; in that, procedures were rewritten, technicians retrained, equipment
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deficiencies corrected and the system was repeatedly tested. With the exception of the dissolved gas issue, the sample results met the appropriate acceptance criteri Surveillance tests are to be conducted on a quarterly basis for routine system maintenance
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and an EP PASS drill will be conducted annually according to the E-Plan and NUREG 073 The licensee is continuing to assess their method for retrieving and analyzing a dissolved gas sample. The licensee's requested change to the PASS commitment concerning total dissolved gas analysis accuracy for lower concentrations is currently under review by the NRC Office of NRR. Until this issue is resolved, NRC violation 50-423/98-01-01 will remain open. However, based upon the demonstrated improvements to the PASS and the reasonable assurance of system functionality, closure of this issue as part of SIL ltem 83 i for Unit 3 (see Section P8.4 below) is deemed appropriat ~
P8.2 (Closed) iFi 50-245,336,473/98-80-02: Licensee corrective actions to internal audit j (92904) !
l l During the inspection period for IR 98-80, the inspectors reviewed a Nuclear Oversight (NOS) audit of the EP program (MP-97-A12-02) which had been completed in December l 1997. The inspectors noted that this audit was detailed and very self-critical. Based 1 l upon the assessment made by the audit, the inspectors concluded that additional corrective actions were required on the part of the EPSD in order to certify the EP function j as ready for plant restart. The inspectors noted that the EPSD had outlined a corrective i
action plan (CAP) to respond to 162 issues generated from the audit and that some corrective actions had already been completed at that tim At the c!ose of this inspection, the inspectors determined that all but one of 162 issues had been completed by EPSD. NOS had reviewed and verified EPSD's responses to the 162 issues. The inspectors sampled severalissues and verified that the corrective action had been completed and were acceptable. The inspectors also reviewed several corrective action plans and determined them to be satisfactory to address the EP-related condition l
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reports (CRs). NOS piens to conduct a review six months after the CRs are closed to verify the effectiveness of EPSD's corrective action NOS personnel stated that EPSD's threshold for problem identification had been too high in the past. NOS has recently noted an improvement of the EPSD's questioning attitude as demonstrated by the identification of issues and the use of CRs to track and resolve -
issues. The inspectors determined that EPSD had reviewed their own tracking system (EPSTAR) to ensure that no items met the criteria for a CR. EPSD has restricted usage of EPSTAR to preclude tracking significant issues that are to be entered in the site tracking system. The inspectors determined that the EPSD has been more self-critical than in the past as indicated by the increased number of CRs generated since the December 1997 audit. NOS has been evaluating, scoring and trending EPSD's performance in a variaty of areas, such as, drill performance and response to action items. EPSD's performance has been improving since the December 1997 audit. On May 5,1998, NOS issued a memorandum stating that EP was ready for restart pending completion of three issue Since that memorandum, the remaining issues have been resolve The inspectors also reviewed a NOS assessment of the licmsee's April 15,1998 drill in which the response organization's performance was overall satisfactory, but several issues f were identified. The inspectors interviewed NOS personnel, as well as EPSD personnel, regarding the issues and determined that the issues were not indicative of significant programmatic problems. NOS was satisfied with EPSD's ability to identify performance deficiencies during the dril NOS personnel stated the EP program has improved and is in acceptable condition, but also acknowledged that additionalimprovements can be implemented. Based upon NOS's satisfaction with the status of EPSD's responses to audit issues and upon information gathered by the inspectors which indicated that appropriate corrective actions have been performed by the EPSD, inspector followup item IFl 98-80-02 for all three MTstone units is close P8.3 (Closed) VIO 50-245.336.423/97-81-03: Revision 24 of the Emeraency Plan (92904)
In the emergency preparedness inspection report for all three Millstone units, IR 97-81, a violation is documented in which the licensee made revisions to the emergency plan which decreased its effectiveness without prior NRC approval. In some instances, the plan as j changed no longer met the standards of 10 CFR 50.47(b) and the requirements of i Appendix During this inspection period, the inspectors reviewed Revision 24 of the licensee's i emergency plan to determine if the revision corrected concerns identified in NRC Violation l 50-245,336,423/97-81-03. The inspectors determined that the previously identified j l
concerns were resolved with the issuance of Revision 24 of the emergency plan. The plan j provides for the evacuation of site personnel in about 30 minutes and a goal to complete
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accountability of site personnel within 45 minutes. The emergency response organization is clearly described and the licensee's emergency response staffing goals are adequate to i
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meet the established criteria. Items omitted from previous emergency plan revisions have been appropriately addressed in Revision 2 The inspectors concluded that Revision 24 of the Millstone Nuclear Power Station Emergency Plan corrected the previously identified concerns and provides an adequate planning basis for an acceptable state of onsite emergency preparedness in accordance with the standards of 10 CFR 50.47(b), and the requirements of Appendix E to 10 CFR 5 The critaria of NUREG-0654 have been adhered to as described in Appendix G of the Emergency Plan. The emergency action levels are consistent with the methodology outlined in NUMARC/NESP-007. Violation 50-245,336,423/97 81-03is considered close P8.4 Closure of Significant Issues List (SIL) Items (Closed - Unit 1 SIL ltem 43, Unit 2 i SIL ltem 16, Unit 3 SIL ltem 83)
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Based upon satisfactory licensee corrective actior.s to address PASS, Nuclear Oversight
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audit findings, and an approved revision to the emergency plan, as mentioned above, SIL l Items 43,16, and 83 for Units 1,2, and 3 respectively are close {
i S1 Conduct of Security and Safeguards Activities l Inspection Scooe (81700)
Determine whether the conduct of security and safeguards activities meet the licensee's commitments in the NRC-approved security plan (the Plan) and NRC regulatory requirements. The security program was inspected during the period of May 4-8,199 Areas inspected included: access authorization program; alarm stations; communications; protected area (PA) access control of personnel and packages and PA access control of vehicle Observations and Findinas Access Authorization Prooram. The inspectors reviewed implementation of the Access Authorization (AA) program to verify implementation was in accordance with applicable regulatory requirements and Plan commitments. The review included an evaluation of the effectiveness of the AA procedures, as implemented, and an examination of AA records for 10 individuals. Records reviewed included both persons who had been granted and had been denied access. The AA program, as implemented, provided assurance that persons granted unescotted access did not constitute an unreasonable risk to the health and safety of the public. Additionally, the inspectors verified by reviewing access denial records and applicable procedures, that appropriate actions were taken when individuals were denied access or had their access terminated. Those actions included the availability of a
- formalized process that allowed the individuals the right to appeal the licensee's decision.
l Alarm Stations. The inspectors observed operations of the Central Alarm Station (CAS) and the Secondary Alarm Station (SAS) and verified that the alarm stations were equipped with appropriate alarms, surveillance and communications capabilities. Interviews with the alarm station operators found them knowledgeable of their duties and responsibilities. The
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inspectors also verified, through observations and interviews, that the alarm stations were l continuously manned, independent and diverse so that no single act could remove the i plant's capability for detecting a threat and calling for assistance, and the alarm stations ( did not contain any operational activities that could interfere with the execution of the l detection, assessment and response function Communications. The inspectors verified, by document reviews and discussions with
! alarm station operators, that the alarm stations were capable of maintaining continuous l intercommunications, communications with each security force member (SFM) on duty,
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and were exercising communication methods with the locallaw enforcement agencies as l committed to in the Pla PA Access Control of Personnel and Hand-Carried Packaaes. On May 5,1998,the !
inspectors observed personnel and package search activities at the personnel access portals. The inspectors determined, by observations, that positive controls were in place to ensure only authorized individuals were granted access to the PA and that all personnel and hand carried items entering the PA were properly searche PA Access Control of Vehicles. On May 7,1998, the inspectors observed vehicle access-control activities at the main vehicle access control entry point. The observations included SFM's verification of vehicle authorization and escort requirements and the performance of vehicle searches prior to granting PA access. The inspectors concluded that vehicles were f being controlled and maintained in accordance with the Plan and applicable procedure ' Conclusions The licensee conducted its security and safeguards activities in a manner that protected public health and safety. This portion of the program, as implemented, met the licensee's commitments and NRC requirement S2 Status of Security Facilities and Equipment Inspection Scope (81700)
Areas inspected were: PA assessment aids and personnel search equipmen Observations and Findinas PA Assessment Aids. On May 6,1998, the inspectors evaluated the effectiveness of the assessment aids, by observing on closed circuit television, a security force member (SFM)
i conducting a walkdown of the PA. The assessment aids, in general, had good picture quality and good zone overlap. Additionally, to ensure Plan commitments are satisfied, the
- licensee has procedures in place requiring the implementation of compensatory measures in i the event the alarm station operator is unable to properly assess the cause of an alar Personnel and Packaae Search Eauipment. On May 5,1998, the inspectors observed both the routine use and the daily performance testing of the licensee's personnel and package search equipment. The inspectors determined, by observations and procedural reviews,
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! Conclusions The licensee's security facilities and equipment were determined to be well maintained and reliable and were able to meet the licensee's commitments and NRC requirement S3 Security and Safeguards Procedures and Documentation Insoection Scooe (81700)
Areas inspected were: implementing procedures and security event log Observations and Findinas Security Proaram Procedures. The inspectors verified that the security procedures were consistent with the Security Plan commitments, and were properly implemented. The verification was accomplished by reviewing selected implementing procedures associated with PA access control of personnel, PA access control of vehicles, and testing and maintenance of personnel search equipmen Security Event Loos. The inspectors reviewed the Security Event Log for the previous
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twelve months. Based on this review, and discussion with security management,it was 1 determined that the licensee appropriately analyzed, tracked, resolved and documented j safeguards events that the licensee determined did not require a report to the NRC within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Additionally, the inspectors noted, during the review of the safeguards event logs, that since the last inspection conducted in February 1997, there was a significant reduction in log entries associated with personnel error Conclusions Security and safeguards procedures and documentation were being properly implemente Event Logs were being properly maintained and effectively used to analyze, track, and resolve safeguards event S4 Security and Safeguards Staff Knowledge and Performance
' Inspection Scope (81700)
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Area inspected was security staff requisite knowledg l Observations and Findinas f Security Force Reauisite Knowledae. The inspectors observed a number of SFM's in the performance of their routine duties. These observations included alarm station operations, l
j personnel and package searches, and vehiclo searches. Additionally, the inspectors l
interviewed SFMs and, based on the responses to the inspectors, determined that the -
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SFMs were knowledgeable of their responsibilities and duties, and could effectively carry out their assignments, Conclusions The SFMs adequately demonstrated that they have the requisite knowledge necessary to effectively implement the duties and responsibilities associated with their positio S5 Security and Safeguards Staff Training and Qualification Inspection Scoce (81700) l
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Areas inspected were security training and qualifications record ! Observations and Findinos j i
Security Trainina and Qualifications. On May 6,1998, the inspectors randomly selected and reviewed security training and qualification records of 11 security force members .
Physical and requalification records were inspected for armed and supervisory personne The results of the review indicated that the security force was being trained in accordance with the approved security training and qualification pla l On May 4,1998, the inspectors observed initial qualification classroom training. The inspectors determined, based on the observations and discussions with security supervision, that security training instructors do not receive formalized instructor training prior to assuming classroom instructor responsibilities. Northeast Utilities (NU) does have a Basic Instructor Certification Program but security instructors have not been afforded the opportunity to take the training. The inspectors also noted that during the classroom training observations, no training aids were used. The inspectors were informed that due I to limited classroom space there was not enough room to store the training aids. The inspectors discussed the importance of effective training with recurity supervision and were informed that to enhance the effectiveness of the security training program, NU training department willinclude security in the next Basic Instructor Certification Progra Additionally, the licensee stated that an effort would be made to acquire additional clastroom space so that the training aids could be used to enhance the effectiveness of the instructor's presentation Trainina Records. The inspectors were able to verify, by reviewing training records, that the records were properly maintained, accurate and reflected the current qualifications of the security force staff . Conclusions Security force personnel were being trained in accordance with the requirements of the Plan. Training documentation was properly maintained and accurate and the training provided by the training staff was adequate.
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S7 Quality Assurance in Security and Safeguards Activities Inspection Scoce (81700)
Areas inspected were: audits, problem analyses, corrective actions and effectiveness of management control Observations and Findinas Audits. The NRC inspectors reviewed the 1997 QA audit of the security program, conducted September 15 - 29,1997, (Audit No. MP-97-A09-02) and the 1997 QA audit of the fitness-for-duty (FFD) program, conducted May 5 - 16,1997, (Audit No. MP 97-A05-01). The audits were condacted in accordance with the Plan end FFD rule. To enhance the effectiveness of the audits, both audit teams included en independent technical specialist. The security audit report identified no findings and six programmatic deficiencies. Three deficiencies addressed vehicle and personnel access controlissues, two deficiencies were associated with the testing and maintenance of security equipment, and one deficiency was associated with portions of the vehicle barrier system not being properly described in the Plan. The FFD audit identified one finding and five programmatic deficiencies. The FFD finding was associated with the administration of supervisory FFD training. Two deficiencies were associated with procedural adherence, and three deficiencies were associated with administrative issues. The inspectors determined that the findings were not indicative of programmatic weaknesses, and the findings would enhance program effectiveness. Inspectors' discussions with security management and FFD staff revealed that the responses to the findings were completed, and the corrective actions were effectiv Problem Analyses. The inupectors reviewed data derived from the security departrnent's self-assessment program. Potential weaknesses were being properly identified, tracked, and trende Corregtive Actions. The inspectors reviewed corrective actions implemented by the licensee in response to the QA audits and self-assessment program. The corrective actions were effective, as evidenced by a reduction in personnel performance issues and logable safeguards event Effectiveness of Manaaement qgntrols. The inspectors observed that the licensee had programs in place for identifying, analyzing and resolving problems. They include the performance of annual QA audits, a departmental self-assessment program and the use of industry data such as violations of regulatory requirements identified by the NRC at other
- facilities, as a criterion for self-assessment.
l During the previous inspection, the inspectors found that the controls for identifying, resolving, and preventing programmatic problems wue not always effective. To aodress the concerns, the licensee revised the self-assessment program and incorporated additional procedural guidance. The guidance requires additional tracking and trending of identified weaknesses and the issuance of a condition report, that can not be closed, until the corrective actions have been implemented. The self-assessment program was well
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defined, structured, and since January 1,1998, security supervision has conducted 10 [
self-assessment j Conclusions i
The review of the licensee's audit program indicated that the audits were comprehensive in l scope and depth, that the audit findings were reported to the appropriate level of management, and that the program was being properly administered. In addition, a review l of the documentation applicable to the self-assessment program indicated that the program I was being effectively implemented to identify and resolve potential weaknes V. Manaaement Meetings
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X1 Exit Meeting Summary
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The inspectors presented the inspection results to members of licensec management at i separate meetings in each unit at the conclusion of the inspection. The licensee l acknowl edged the findings presente X1.1 Final Safety Analysis Report Revim While perforrning the inspections which are discussed in this report, the inspectors reviewed the applicable portions of the UFSAR that related to the areas inspecte Inconsistencies have been noted between the wording of the UFSAR and the plant ;
practices, procedures and/or parameters observed by the inspectors as is documented in Sections U3.07.2, U3,M3.1, and P8.1.
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o5 INSPECTION /ROCEDURES USED i
! IP 37001 10 CFR 50.59 Safety Evaluation Program IP 37550 Engineering
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IP 37551 Onsite Engineering IP 40500 Effectiveness of Licensee Controls in identifying, Resolving, and Preventing Problems l
lP 61726 Surveillance Observations IP 62707 Maintenance Observations IP 71707 Plant Operations IP 81700 Physical Security Program for Power Reactors ,
l IP 83729 Occupational Exposure During Extended Outages i
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IP 92700 Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities IP 02901 Followup - Plant Operations IP 92902 Followup - Ma.intenance IP 92903 Followup - Engineering
- - IP 92904 Followup - Plant Support Tl 2515/109 Inspection Requirements for Generic Letter 89-10, Safety Related Motor-Operated Valve Testing and Surveillance
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ITF.MS OPENED, CLOSED, AND DISCUSSED Opened TYPE DOCKET NUMBER DESCRIPTION SECTION URI 50-245 98-208-01 GENERIC IMPLICATIONS - Indicator VALVE U 1.M RECYCLING URI 50-336 98-208-02 PROCEDURALLY IMPLEMENTED U2 E TEMPORARY MODIFICATIONS l NCV 50-336 98-208-03 BYPASSED REFUELING MACHINE U2.E OVERLOAD CUTOFF VIO 50-423 98-208-04 FAILURE TO IMPLEMENT PLANT HEATUP U3.0 {
PROCEDURE i
IFl 50-423 98-208-05 MEPL ITEMS 1-6 U3.M )
I IFl 50-423 98 208-06 COMPI ETION OF MOV PROGRAM CLOSURE U3.E I ISSUES -
IFl 50-423 98-208-07 CONTAINMENT RSS CUBICLE FLOOD U3.E POTENTIAL I
Cloped eel 50-336/96-06-05 SECTION U2.0 NCV 50-336/98-208-03 SECTION U2.E {
URI 50-423/96-01-07 SECTION U3.0 IFl 50-423/95-07-10 SECTION U3.M IFl 50-423/98-82-01/02/06/07 SECTION U3.E.1. '
IFl 50-423/98-82-03/04/05/08/09 SECTION U3.E VIO 50-423/98-82-10 SECTION U3.E ]
URI 50-423/98-C 2-11 SECTION U3.E J IFl 50/245/336/423/98-80-02 SECTION IV.P VIO 50-245/336/423/97 81-03 SECTION IV.P VIOs (Ref: December 10,1997 NOV/CP Letter) SECTION U3.0 including: l VIO eel 96-201-01 SECTION U3.07.2 i
VIO eel 96-201-43 SECTION U3.M VIO Eels 96-201-02/04/05/06/07/08 SECTION U3.E Discussed URI 50-245/97-02-02 SECTION U1.M2.1
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The followina LERs were also closed durina this inSDeClion:
50-245/96-044-00/01 SECTION U1.M /96-035-00/01l02 SECTION U1.E /96-013 SECTION U1.E /96-033-00 SECTION U2.E /97-017-00 SECTION U2 E /97-030-00 SECTION U2.E /97 010-00 SECTION U3.M /97-011-00 SECTION U3.M /97-061-00 SECTION U3.M /98-006-00 SECTION U3.E The followina LER was discussed duti.m this insDection:
50-423/97-046-00 SECTION U3.E )
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i 98 LIST OF ACRONYMS USED ACR(s) adverse condition report (s)
AFW auxiliary feedwater AITTS action item tracking and trending system AWO(s) automated work order (s)
BOM bill of materials CAP (s) corrective action plan (s)
CAS central alarm station CCPL chemical consumable product list CFR Code of Federal Regulations CGD commercial grade dedication i CR(s) condition report (s)
l CTL Construction Technology Laboratories DBE design basis earthquake DCN(s) design change notice (s)
l DCR design change record DTE dynamic test evaluation l ECP Employee Concerns Program EDG(s) emergency diesel generator (s)
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EDO Executive Director of Operations eel (s) escalated enforcement item (s)
EOP(s) emergency operation procedure (s)
EPRI Electric Power Research Institute ERT event review team ESF engineered safety feature ESFAS emergency safety features actuation system FFD Fitness for Duty FSAR Final Safety Analysis Report FSARCR(s) Final Safety Analysis Report Change Request (s)
GL Generic Letter gpm gallons per minute i GTG gas turbine generator HAC high alumine cement J
HELB high energy line break HPSI high pressure safety injection ICAVP Independent Corrective Action Verification Program IFl inspector follow item
- IGSCC intergranular stress-corrosion cracking IPTE infrequently performed test IR(s) Inspection Reports (s)
IRT independent review team ISI inservice inspection KSREL key safety-related equipment list LER(s) licensee event report (s)
l LHC Little Harbor Consultants i LOCA loss of coolant accident i LPCI low pressure coolant injection
LPSI low pressure safety injection MDAFW motor driven auxiliary feed water MEPL(s) material, equipment, and parts list (s)
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MFW main feedwater MIF(s) materialissue forms (s)
MIMS materials information management system MOV(s) motor-operated valve (s)
MR(s) material request (s)
MSARBV(s) main steam atmospheric relief bypass valve (s)
l MSLB main steam line break NCR(s) nonconformance report (s)
NCV non-cited violation -
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- NGP(s) nuclear guidance procedure (s)
NNECO Northeast Nuclear Energy Company NOV(s) Notice of Violation (s)
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NRR Nuclear Reactor Regulation NSAB nuclear safety assessment board NSR nonsafety-related OD(s) operability determination (s)
OP(s) operating procedure (s)
OPC ordinary Portland cement OSTI operational safety team inspection P&lD piping & instrumentation diagrams PDR Public Document Room PMMS production maintenance management system PORC plant operation review committee PORV(s) power operated relief valve (s)
PPM performance prediction model l PPP performance prediction program QA quality assurance QAS Quality and Assessment Services RBCCW reactor building closed cooling water 4 RCS reactor coolant system RFO refueling outage RHR residual heat removal RSS recirculation spray system RWST refueling water storage tank SCWE safety conscious work environment SER(s) safety evaluation report (s)
SFM(s) security force member (s)
l SGTR steam generator tube rupture l SlAS safety injection actuation signal ,
SIL significant item list l SORC site operations review committee SPROC special procedure SSC(s) structures, systems, and component (s)
SSER supplemental safety evaluation report TBN total base number l
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.100 TDAFW
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- turbine driven auxiliary feedwater TMl Three Mile Island -
l TQR training qualification record TR(s) trouble report (s)
TS(s) technical specification (s)
UFSAR updated final safety analysis report URl(s) unresolved item (s)
USQ(s) unresolved safety question (s)
VETIP ' vendor evaluation technical information program
. VIO ' violation l = VTM(s) vendor technical manual (s)
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