IR 05000336/1998201
| ML20237D310 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/12/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20237D300 | List: |
| References | |
| 50-336-98-201, NUDOCS 9808260016 | |
| Download: ML20237D310 (42) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Report No.: 50-336/98-201 l l Docket No.: 50-336 , License No.: DPR-65 Licensee: Northeast Nuclear Energy Company l Facility: Millstone Unit 2 I l Location: Millstone Nuclear Power Statio Rope Ferry Road L Waterford, Connecticut 06385 Dates: April 13 through May 8,1998 Inspectors: Richard P. McIntyre, ICAVP, Leader, Team 2B Special Projects Office (SPO) Brian Hughes, Operations inspector, SPO Donald Prevatte, Mechanical Engineer, Contractor * Robert Quirk, I&C Engineer, Contractor * , Omar Mazzoni, Electrical Engineer, Contractor * ! Victor Ferrarini, Civil / Structural Engineer, Contractor *
* Contractors from Parameter, In Approved by: Peter S. K'oltay, Branch Chief Special Projects Office Office of Nuclear Reactor Regulation l
Enclosure 2 _ , 9909260016 990812 ! PDR ADOCK 05000336 e PM _ _ _ _ _ _ _ _ _ _ _ _ _ . _
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SUMMARY On April 13 through May 8,1998, a team from the U.S. Nuclear Regulatory Commission's (NRC's) Special Projects Office, Office of Nuclear Reactor Regulation, in accordance with the guidelines outlined in SECY-97-003, " Millstone Restart Review Process," conducted a Tier 3 inspection at Millstone Unit 2 and at the offices of Parsons Power Group inc. (Parsons), the Unit 2 Independent Corrective Action Verification Program (ICAVP) contracto The purpose of this Tier 3 inspection was to independently assess the licensee's ability to identify and resolve licensing-basis deficiencies focusing but not limited to a period of the Configuration Management Plan (CMP) implementation; determine if the licensee's change processes were adequate to maintain the Unit 2 design and licensing bases and to assess the effectiveness of the Tier 3 aspects of Parson's ICAVP. The review evaluated a sample of changes made to the facility configuration since issuance of the operating license and a review of the processes that govemed those changes. The Millstone Unit 3, Tier 3, inspection reviewed and evaluated the change processes in a forward looking manner and determined that many change process improvements, both programmatically and procedurally, had been implemented and such processes can assure an effective configuration managemen The team's independent review addressed a large number of change processes. The combined sample size was sufficiently large to provide overall indications of the licensee's approach to work, problem solving skills, technical abilities, and commitment to quality. The
. team evaluated both the licensee's past and present performance during the Tier 3 inspection, in addition to Parson's implementation of the ICAVP. The team reviewed approximately 160 changes to the plant implemented since 1990. In addition, the team selected 50 past changes for which the contractor had completed its review to facilitate the review of the contractor's ICAVP. A list of plant changes reviewed by the team is contained in an attachment to this repor Based on indications from this overall sample. The team determined that the existing Millstone Unit 2 change processes and procedures reviewed by the team met the requirements of 10 CFR Part 50, Appendix B, " Quality Assurance," and if adequately implemented, should maintain the Unit 2 design and licensing basis. The team found that the licensee generally used good engineering practices, however technical work often lacked appropriate attention to
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detail. To achieve adequate implementation, actions such as increased attention to detail are necessary and the depth and breadth of problem resolution must improve to get beyond the types of findings identified during this inspectio l
- As part of Northeast Nuclear Energy Company's (NNECO's) review of the final results of this inspection, in combination with the results of the Parsons ICAVP effort, NNECO should {
evaluate the need to enhance the implementation of its current change processes.
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The team found that the Parsons Tier 3 ICAVP review was conducted in accordance with the NRC-approved audit plan and project procedures and that the reviews were conducted in a thorough and critical manner. Generally, the findings identified by the team were consistent with the findings identified by the Parsons ICAVP. For example, Parsons identified similar problems with the threshold for writing SEs for FSARCRs.
The team identified eight violations during the site inspections. Some of the violations had multiple examples. The eight violations are considered as ICAVP significance Level 3 findings.
The licensee had identified and was in the process of correcting many system and process problems from its Unit 2 CMP review. In accordance with NRC policy, when the team identified a problem that the CMP had already identified and had or was in the process of correcting, the team did not issue a violation.
Similar to previous NRC inspection teams' conclusions, the team recommended that the licensee give continued attention to lowering the threshold for performing safety evaluations in accordance with 10 CFR 50.59.
Additionally, the team made a number of observations of strengths and good practices. For example, while the team found instances of inadequacies for past work, the team observed a current positive attitude and a major improvement in the approach to solving plant problems.
The team noticed a good level of response to the team's questions raised during the inspection, evidencing relatively good information retrieval capabilities, as well as adequate technical knowledge in providing backup evaluations that were missing in the reviewed documentation.
The team also concluded that the licensee had identified and resolved many important CMP deficiencies and was continuing the process of implementing several improvements to program weaknesses. Although the team had findings, the number of findings was not unusual for this type of inspectio ! l l
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' Backaround On August 14,1996, the U.S. Nuclear Regulatory Commission (NRC) issued a Confirmatory Order (Order) to the Northeast Nuclear Energy Company (NNECO or the licensee) requiring
) completion of an Independent Corrective Action Verification Program (ICAVP) before the restart of any Millstone unit. The Order directed the licensee to obtain the services of an organization independent of the licensee and each facility's design contractor to conduct a multi-disciplinary review of Millstone Units 1,2, and 3. The staff approved NNECO's selection of the Parsons Power Group Inc (Parsons) to perform the ICAVP for Millstone Unit 2 on July 15,199 The Order also stated that an ICAVP audit plan was to be developed by the independent organization and that the audit plan must describe the following information:
(1) the conduct of an in-depth review of selected systems' design and licensing bases (2) risk- and safety-based criteria for selection of systems for review (3) activities to ensure that the quality of results of the licensee's problem identification and corrective action programs on the selected systems is representative of and consistent with other systems (4) procedures and schedules for parallel reporting of findings of the ICAVP team to both the NRC and the licensee (5) proceuures for the ICAVP team to comment on the licensee's proposed resolution of the ICAVP team's findings and recommendations The Order further stated that the scope of the ICAVP shallinclude the following activities: (1) a review of engineering design and configuration control processes (2) verification of current, as modified, conditions against design- and licensing- basis documentation (3) verification that the design- and licensing-basis requirements have been properly translated into operating procedures and maintenance and test procedures (4) verification of system performance through rev;ew of specific test records or observation of selected testing of particular systems (5) review of proposed and implemented corrective actions for design deficiencies identified by the licensee
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. The Three Tier Process in a paper to the Commission, SECY-97-003, " Millstone Restart Review Process," dated January 3,1997, the staff described the Millstone restart review process. To provide necessary assurance to support a unit restart decision, the staff's expectation, described in SECY-97-003, was that the ICAVP would encompass the aspects of configuration control described in a Tier 3 approac In Tier 1, systems were selected to test the thoroughness of the licensee's reviews in identifying potential nonconformance with the design and licensing bases. The systems selected were the high pressure safety injection (HPSI) system and the refueling water storage tank (RWST) as one system; the auxiliary feedwater (AFW) system and the condensate storage tank (CST) as one system; emergency diesel generator (EDG) and support systems ( which includes six systems); and the radiological release control systems (which includes three systems). The ICAVP team was tasked to conduct a thorough review of all design changes made to these systems after the issuance of the operating license, the remaining part of the original system configuration, and all operational aspects of these systems, including maintenance, surveillance testing, and training. Parsons was also expected to review the licensee's corrective actions for previously identified design-related deficiencies for the selected system The Tier 2 objective was to verify that critical design characteristics of systems relied upon to mitigate the consequences of accidents analyzed in the Final Safety Analysis Report (FSAR)
were consistent with those used in the design of the mitigation system and the accident analyses. The scope of the ICAVP Tier 2 review was focused on verifying critical design characteristics; therefore, the Tier 2 review involved more systems than the Tier 1 review, The Tier 3 objective was to provide insights into the effectiveness of the various change processes in controlling the plant's configuration over the lifetime of the plant. The review ' evaluated a sample of changes made to the facility's configuration since the issuance of the operating license and a review of the processes that governed those changes. This included processes for calculation changes, proposed Technical Specification (TS) changes, temporary modifications, drawing changes, procedure changes, setpoint change requests, and replacement item evaluation .2 Configuration Manaaement Chance Process (Tier 3) The NRC configuration management change process team inspection (Tier 3) was aimed at determining the effectiveness of Miilstone's CMP and Parsons'ICAVP reviews. This was accomplished in two ways, (1) reviewing independent samples not rev,'ewed by Parsons to form independent conclusions regarding the licensee's work, and (2) reviewing of selected samples also reviewed by Parsons and comparing the team's observations and findings with those identified by Parsons.
The independent samples represented the licensee's work over the life of the plant, and the sample reviews of both Parsons and the team were focused on the quality of the change
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processes as well as on the technical work being accomplished. The team's independent sample review, however, was more heavily weighted toward plant modifications completed since 199 To accomplish its objectives, the team evaluated the licensee's procedures that control changes to plant calculations, TSs, permanent and temporary modifications, drawings, set points, replacement items, the Final Safety Analysis Report (FSAR), computer software, and operations, maintenance, and surveillance procedures. The team inspected approximately 160 changes made to Unit 2 since January 1990. The effectiveness of Parsons's review was verified by inspecting approximately 50 of the past changes reviewed by Parsons. The specific changes and procedures inspected by the team are listed in Appendix The team focused on verifying the following: (1) the licensee appropriately considered that design disciplines were to be included in the review of the design change, (2) the design change was appropriately reviewed by plant management, (3) the design package was properly evaluated to determine if the change constituted an unreviewed safety question, (4) the appropriate design bases were utilized in the review, (5) the change used properly qualified components, (6) the design change was properly tested after installation, (7) the operators received the appropriate training on the change, (8) the operating procedures were updated to incorporate the change, and (9) the system interaction was considered in the change proces The team reviewed the following significant engineering control processes used to maintain the Unit 2 design and licensing bases: e Engineering / Licensing Documents:
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setpoint changes
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specification revisions 3
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drawing revisions
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calculation revisions
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licensing c'nument changes
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Use-as-is nonconformance reports (NCRs)
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engineering work request (EWR) e Parts Dedication, Substitution, and Safety Classification
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commercial grade dedication
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material equipment parts list (MEPL) safety classification changes s Operations and Maintenance
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operation procedures ! - abnormal operating procedures
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surveillance procedures
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maintenance procedures
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inservice inspection (ISI) procedures
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inservice testing (IST) procedures
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American Society of Mechanical Engineers (ASME) Section XI, Repairs, Replacements, and Modifications
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temporary changes / jumper, lifted leads, and bypasses The team selected a sample of changes to Unit 2 systems, structures, or components for review that represented all of the above listed change processes. Samples chosen were reviewed by the team in the following disciplines: mechanical systems, electrical systems, instrumentation and control systems, structures, piping and pipe supports, and plant operation .0 Mechanical Scone of Review The team evaluated the mechanical systems design change processes by reviewing the following documents and change packages not reviewed by Parsons: ' e 13 Plant Design Change Request modification packages (PDCRs) e 1EWR e 5 Plant Technical Specification Change Requests (PTSCRs) e 4 nonconformance reports (NCRs) e 2 calculations e 4 procedure The team reviewed the following documents and change packages that were also reviewed by the Parsons ICAVP: e 4 PDCRs e 6 EWRs e 2 NCRs e 3 FSAR Change Requests (FSARCRs) e 2 temporary modifications e 4 commercial grade dedications e 2 station procedure changes e 2 MEPL changes i Observations and Findinas 2. Independent Site Findings
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2.2. Failure to Leak Test Containment Sump Isolation Valves Modification package PDCR 2-052-95 was generated to plug a hole in the inboard disk of the HPSI containment sump suction valves,2-CS-16.1 A&B. These valves were double-disk, gate-type valves, and holes had been drilled under a previous modification, PDCR 2-023-95, that
addressed pressure locking concems. However, this earlier modification had proven
: unsatisfactory in that it allowed leakage of the RWST into the containment sum The team observed that neither of these modification packages had specified post-modification leakage testing for these valves even though they had been disassembled. Additionally, the team discovered that the valves were not included in any periodic leakage testing program, either under 10 CFR Part 50, Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," or the IST requirements of ASME Code, Section XI, as required by 10 CFR 50.55a, " Codes and Standards."
The licensee maintained that such tests were not required for the following reasons:
(1) Since the valves were in the Appendix J, Type A, integrated leak rate test boundaries, they were not required to be Type C tested. This was not valid, first, because Type A <
testing methodology would not reveal the leakage characteristics of these individual { valves, and second, because Appendix J did not allow Type A tests to be in lieu of Type C tests, but rather as additions to Type C test l
- (2) FSAR Table 5.2-11, " Containment Structure Isolation Valve Information," listed these l l valves as not requiring Type C leakage testing because they were categorized as
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"special." In response to the team's inquiries as to how they were "special," the
! licensee stated that since they were required to be open post-loss-of-coolant accident (LOCA) to provide the accident recirculation phase flow paths for emergency core cooling systems (ECCS) and containment cooling, they were not required to be teste However, this was only one of their tag safety functions. Their second safety function ! provides containment isolation for passive failure of an HPSI system component, such as a pump seal. For this case, the allowable leakage should have been some fraction of the design basis of 12 gallons per hour (gph) total ECCS leakage reflected in the l accident radiological dose analyses. Therefore, these valves were specifically required ! to be Type C tested per 10 CFR Part 50, Appendix J, Paragraph II.H.3, which includes l valves "... required to operate intermittently under postaccident conditions...," as these l valves would be for the above descrl bed single failure scenario.
l Testing was also required by the ASME Section XI, IST leak testing program, as stated by 10 CFR 50.55a, " Codes and Standaids", Section (f), " Inservice Testing Requirements." Per Article IWV-2200(a) of Section XI, such valves were classified as Category A valves, " valves for which seat leakage is limited to a specified maximum amount in the closed position of fulfillment of their function," by virtue of the 12 gph ECCS leakage limit. Article IWV-3421 required that such " Category A valves shall be leak tested "...in a manner that demonstrates functionally adequate seat tightness..."(i.e., at a rate that would not cause the design basis ECCS leakage 1 limit to be exceeded).
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i i At the conclusion of the inspection, the licensee indicated in its inspection Request No. 048 response, that "The Third 10 Year IST Program will be revised to include leak testing 2-CS-16.1 A/B to periodically verify the containment sump line can be isolated and limit leakage to less than calculated offsite dose limits during a SRAS event coincident with a passive failure in the ECCS system."
10 CFR Part 50, Appendix B, Criterion XI, " Test Control," requires that, "A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in the applicable design documents."
The failure to include valves 2-CS-16.1 A&B, which have a safety-related leakage limiting design function on the basis of the offsite and control room accident dose calculations, in the licensee's leakage testing programs is considered a violation of 10 CFR 50, Appendix B, Criterion XI,
" Test Control," 10 CFR Part 50, Appendix J for Type C valve leakage, and 10 CFR 50.55a, which requires inservice testing for valve leakage per ASME Code, Section X This was an ICAVP significance Level 3 finding. (VlO 50-336/98-201-01, Example 1)
2.2. Failure to Leak Test Refueling Water Storage Tank (RWST) Suction Valves RWST suction valves 2-CS-14A&B and 2-CS-13.1 A&B were the ECCS suction check valves and motor-operated isolation valves, respectively, from the RWST. In a LOCA, these valves would be open during the injection phase, and are required to close during the recirculation phase to prevent back leakage of accident water from the ECCS system into the RWST. Such leakage could cause a ground-level radioactivity release from the RWST open ven CFR 50.55a, " Codes and Standards," Section (f), " Inservice TeLing Requirements," requires that such valves be included in the ASME Section XI IST leak testing progra Section XI, Article IWV-2200(a), classified such valves as Category A valves (i.e., " valves for which seat leakage is limited to a specified maximum amount in the closed position of fulfillment of their function." Paragraph IWV-3421 required that such "Ca% gory A valves shall be leak tested...in a manner that demonstrates functionally adequate seat tightness...," (i.e., at a rate less than that which would cause the design-basis offsite or control room accident dose limits to be exceeded).
10 CFR Part 50, Appendix B, Criterion XI, " Test Control," requires that, "A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in the applicable design documents."
Adelitionally, testing was also required by the following two documents:
(1) NUREG 0737, Section Ill.D1.1, " Integrity of Systems Outside Containment Likely to Contain Radioactive Material for Pressurized-Water Reactors and Boiling-Water
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systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels." In a letter to the NRC . dated December 31,1979, the licensee responded by stating that, "Since April 23, 1979, NNECO has implemented an in-service inspection and testing program at Millstone Unit No. 2 as defined in the Summer,1975 Addenda to the ASME Code, l Section XI,'.' to "... provide additional assurance of system and component leak tightness."
(2) Millstone Unit 2, Technical Specification 6.13, " Systems Integrity," also requires that,
"The licensee shallimplement a program to reduce leakage from systems outside containment that would, or could, contain highly radioactive fluids during a serious . transient, or accident, to as low as practical levels."
The failure to include ECCS RWST suction valves 2-CS-14A&B and 2-CS-13.1 A&B in the IST leakage testing program is considered a violation of 10 CFR 50.55a and ASME Section XI, Subsection IWV, for valves for which leakage is limited in order to fulfill of their design functio This was an ICAVP significance Level 3 finding. (VIO 50-336/98-201-01, Example 2) The team also discovered another related discrepancy. Four-inch manual gate valve 2-CS-28 was shown on P&lD 26015, Sheet 2. "High Pressure Safety injection Pumps," and in Station Procedure SP 2606D, Rev.13, " Containment Spray System Alignment, Facility 2," as normally open. For single failure of check valve 2-CS-14B to close, this position would bypass RWST motor-operated suction valve 2-CS-2-CS-13.18, which is required to be closed by the Operator at the beginning of the accident recirculation phase by EOP 2532, " Loss of Primary Coolant."
This would provide a release path to the environment through the RWST atmospheric vent as well as a path for substantial loss. of reactor coolant inventory not accounted for in the accident analyses. It should also be noted that this topic of concern was brought to the attention of licensees in 1991 through information Notice 91 56, " Potential Radioactive Leakage to Tank Vented to Atmosphere."
The team attempted to determine if the licensee correctly translated the design basis requirement for limiting the release of radiation as a result of an accident into the correct i.
' position for valve 2-CS-28 as shown on the HPSI system P&lD, drawing No. 26015, Sheet 2, Rev.15, and into the HPSI system operating procedure, SP 2606, Rev.15, valve lineup. It should be noted that in 1997, the licensee identified the nonperformance of leakage testing of various ECCS system boundaries outside containment as a concern in CR M2-98-0388; however, the above identified valves were not included in the boundaries identified. The team L acknowledged that Millstone Unit 2 had been recently reviewing RWST backleakage through various flow paths, as described in memorandum TS2-98-0122, dated April 9,1998. Therefore, this issue is identified as a inspector Followup Item. (IFl 50-336/98-201-09) l 2.2.1.3_ Service Water Pump Degradation
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NCR 291-164 identified wastage in the 316L stainless steel material (approximately 0.100 inches deep) inservice water pump P5C's column at the edge of a Belzona coating that had-7 l
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' been applied to address a previous material degradation concern. The identified corrective actions were to calculate the maximum allowable degradation, blend the degradation, and coat it with more Belzon I' The initially identified cause for the degradation was erosion / corrosion. However, the team observed that this was inconsistent with the actual conditions, (i.e., low fluid velocities, single phase flow, and 316L stainless steel material, a material typically used to prevent erosion / corrosion). The correct cause of the degradation was not identifie In response to additional degradation observed at the Belzona edge, PDCR 2-208-92 was later issued to coat the remaining area of the pump column, with the exception of the bearing support spider and the column rabbit fits, with Belzona. This document identified two additional degradation causes, microbiological attack and crevice corrosion, but no basis was provided for either diagnosis, and no actions were taken to prevent recurrence, in leaving the bearing support spider and the column rabbit fits uncoated, this PDCR potentially subjected these areas to the same attack mechanism. Failure in either of these aren could cause catastrophic pump failure as a result of a loss of support for the pumps' major component The team also discovered that the cathodic protection system in the intake bay had a history of improperly functioning. This coupled with the observation that the wastage occurred only at the Belzona edge, suggested that the cause might be galvanic action between the stainless steel and the Belzona. The team also noted that the licensee's Belzona application procedure, MP 2721Z, contained no precautions regarding material compatibility between the Belzona and the components to be coate The licensee's failure to identify correctly the root cause of the degradation of the service water pumps and to take appropriate corrective actions to preclude repetition does not meet the requirements of 10 CFR 50, Appendix B, Criterion XVI, " Corrective Action." In response to this finding, the licensee generated condition report (CR) M2-98-1076 and committed to inspect all of the service water pumps before the Unit 2 restar This was an ICAVP significance Level 3 finding. (VIO 50-336/98-201-02, Example 1) 2.2. Improper Reactor Building Close-cooling We9r (RBCCW) Heat Exchanger Relief Valve Setpoint EWR M295316 changed the setpoint of the RBCCW heat exchanger shell side relief valves, 2-RB-303A/B/C, from 150 psig to 165 psig in order to address a problem with the valves lifting during system transients. The heat exchanger's design pressure was 150 psi ASME Code Section Vill, Article UG-134(a), stated, "When a single pressure relieving device is used, it shall be set to operate at a pressure not exceeding the maximum allowable working pressure of the vessel (the design pressure)." By raising the setpoint to 165 psig, the licensee failed to meet this section of the ASME Cod The licensee initially maintained that the 165 psig setpoint was allowed by Article UG-134(b),
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which, for vessels that could be exposed to "an additional hazard," such as fire, allowed a . second relief valve to be used that could be set as high as 110 percent of design pressur However, where only one valve was used, such as with these heat exchangers, even if it was
used to address such an " additional hazard,"its setpoint could not exceed the design pressur CFR Part 50, Appendix B, Criterion lil, " Design Control," requires that, " Measures shall be established to assure that the applicable regulatory requirements and the design basis ...are correctly translated into specifications, drawings, procedures, and instructions." The failure to correctly translate the design requirements of ASME Code Section Vill, Article UG-134(a) for the pressure relief device for unfired pressure vessels into the setpoints of the RBCCW heat exchanger relief valves is considered a violation of the ASME Code and 10 CFR 50, Appendix B, Criterion lil, " Design Control." The licensee acknowledged that the setpoint change had been incorrect and issued CR M2-98-1114 to address this concern.
, This was an ICAVP significance Level 3 finding. (VIO 50-336/98-201-03) 2.2. Qualification of Emergency Core Cooling System (ECCS) and Containment Spray Pump Seals , PDCR 2-147-92 was issued to change the HPSI pump seals from Teflon to Ethylene Propylene ! (also known as EPDM or EPT), and PDCR 2-091-92 was issued to change the seal material in l
' the LPCI pumps because of the low radiation resistance of Teflon. (The containment spray pumps' original equipment seals were already EPDM.) The team asked the licensee to provide documentation to show that the new materials were qualified for the accident environmen Specifically, the team asked the licensee to show that the seal material used was qualified for the combination radiation levels and temperatures to which it would be subjected to during a l LOC In inttempts to demonstrate qualification, the licensee's initial responses addressed many of the material's properties, such as tensile strength, elongation, etc. ' However, the compression set, the most important sealing property, was not addressed. The licensee was asked to show that 3 l the maximum compression set for which the seal would properly function was greater than the j
' compression set that would result from the combination of radiation and temperature that the l seals would experience in an accident. This evaluation was not performe I The team determined that the ECCS pumps' (HPSI pumps P-41 A, B, & C, and LPCI pumps P-42A & B) seal O-ring materials were changed from Teflon to ethylene propylene and post-accident radiation and temperature qualification requirements for this new material were not adequately defined and correctly translated into the seal design. The team acknowledged that Millstone Unit 2 is reviewing all seal qualifications as part of their equipment qualification
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program review and therefore identifies this issue as an inspector Followup Item .
(IFl 50-336/98-201-10)
The team also identified that the "B" HPSI pump seals had been replaced without a DCN
- change for all the corresponding drawings. DCN DM2-S-675-94, which was issued to update the drawings for the "C" HPSI pump seat material change did not adequately update drawing 25203-29168, Sheets 6,10, and 2 l
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l t The failure to make changes to the design of the "B" and "C" HPSI pumps seals without performing the required updates of the associated design drawings is considered a violation of 10 CFR Part 50, Appendix B, Criterion ill, " Design Control." The licensee generated CR M2-98-1094 to document and address this design control concer This was an ICAVP significance Level 3 finding. (VIO 50-336/98-201-04) 2.2. Inadequate Quality Control on RCCW Heat Exchanger Bolting During a plant walkthrough, the team identified that the "C" RBCCW heat exchanger head was bolted on with washers of different sizes and materials. This had previously been identified by the licensee while cleaning the "A" heat exchanger, and a work request to replace the discrepant washers was submitted on February 6,1998. However, contrary to Station Procedure NGP3.05, "Nonconformance Reports, " which requires in Section 1.1 that, "The NCR is used to document and disposition nonconforming materials, parts, components or services...," the licensee generated no NCR to document this discrepancy. Although this observation represented a quality control concern, it did not include any direct safety significanc The licensee failed to adequately control the quality of the buildup of the "C" RBCCW hear exhanger head. Additionally, the licensee failed to properly identify the discrepancies in the bolting of the "C" RBCCW heat exchanger head by issuing an NCR as required by Station Procedure NGP 3.05, "Nonconformance Reports." In response to the team's inquiries, the licensee generated CR M2-98-1108 to document and address this concer This was an ICAVP significance Level 3 finding. (VIO 50-336/98-201-02, Example 2) 2.2. Capability of Motor Control Center Enclosures to Withstand High Energy Line Break Overpressure During a plant walkthrough, the team questioned the structural capability of the enclosures for motor-control centers (MCCS) B51 and BS1 to withstand the overpressure from the most
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severe high energy line break (HELB). For such an event, structural failure of the enclosures could damage the safety-related MCCs which supply power to safe shutdown equipmen At the completion of the inspection, the licensee provided their response to the team's concern in inspection Response No.179, which stated that a calculation change notice (CCN) was being created to tie together the external design pressure issue from a high energy pipe break and the original design calculations. This CCN is being tracked on AR 98008676-0. Therefore, this issue is an Inspection Followup Item. (IFl 50-336/98-201-11) d Conclusions Overall, the review provided indications that the licensee's approach to work problemsolving skills, technical abilities, and commitment to quality to adequate. Some of the technical work lacked appropriate attention to detail, however, the issues identified by the team were oflow safety significanc I
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Parsons planning and documentation for their reviews was rigorous and thorough, and appeared to have the appropriate level of detail to provide the bases for valid appraisals of the licensee's performance of change processes. The depth and attention to detail of the Parsons reviews in the areas of containment leakage testing and inservice appeared to be particularly goo .0 Electrical Scoce of Review The team evaluated the electrical design change processes by reviewing the following documents and change packages not reviewed by the Parsons ICAVP: e . 4 EWRs e 5 jumper device controls e 6 modifications e 1 DCNs e 5 NCRs e 8 PTSCRs e 1 RIE e 1 commerical grade dedication (CGD) The team reviewed the following change packages that were also reviewed by the Parsons ICAVP:
* 1 drawing change and DCN e 1 jumper device control sheet e 1NCR e 1CGD e 1 modification package performed by Parsons under Tier 1 e 1 RIE Observations and Findinos 3. Independent Site Findings 3.2. Replacement of Circuit Breaker Bushings The licensee installed three non-QA bushings in 4.16 Kv safety-related 4.16 switchgear cubicle
, A407, for the "C" service water pump at emergency bus 24D, facility Z2, without performing ( adequate dedication of the non-QA equipment. The acceptance of the non-QA devices was I based on a brief description that considered only a few of the essential characteristics for establishing an equivalency. Consideration was only given to the dimensional characteristics, the weight, and the voltage withstand characteristics. Other critical characteristics that were not included in the analysis of equivalency of devices are as follows:
- _ __-_ - _____ .
e The manufacturer's name and catalog number were not verified for either the non-QA or the QA device. There was no indication given that the same manufacturer and catalog number would apply to both.
l e No information was provided on the current carrying capability for both the QA and non-QA i devices, nor was there any comparison established for these parameters, to ascertain that the non-QA device characteristics were adequate and acceptable for the intended safety-related functio * Licensee testing values for voltage and for length of time exceeded vendor recommended values, without any evidence of an evaluation of the potential for insulation damage due to higher voltage stresses impressed during the tes e There were no acceptance criteria for the Doble testing performed on the device The failure to conduct an adequate replacement item equivalency evaluation and dedication of a non-QA commercial grade item that was installed in a safety-related application is a violation of 10 CFR Part 50, Appendix B, Criterion lil, " Design Control."
This was an ICAVP significance Level 3 findings. (VIO 50-336/98-201-08, Example 1) 3.2. Review of Engineering Evaluation M2-EV-96-0061 The licensee failed to provide any technical evidence to substantiate the basis of Engineering Evaluation M2-EV-96-0061, Rev. O. This was performed in support of DCN No. DM2-00-1466-96. The evaluation incluried no specific reference to the calculation that would support the statement in that the " fault current available over the entire length of the power circuit is adequate to actuate the trip element of any breaker with an instantaneous trip setting up to, and including, the HI setting." Also, the statement that, " Coordination reviews of 480 volt MCC circuits and upstream devices are based upon the largest breaker installed in the MCC," was not referenced to the relevant coordination study. Section 6.0, " References," did not include the coordination study or any calculation No objective technical evidence was evident to support the following engineering evaluation conclusions: e determination that an unreviewed safety question did not exist e no need to perform a safety evaluation e acceptability of the "as found" breaker settings The failure to perform an adequate engineering evaluation for the design change that included objective technical evidence to support the engineering evaluation conclusions is considered a violation of 10 CFR Part 50, Appendix B, Criterion lil, '? Design Control."
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The licensee initiated AR 98008219-02 to track the preparation of a calculation to support the ' engineering evaluatio This was an ICAVP significance Level 3 finding. (VIO 50-336/98-201-05, Example 1) 3.2. Review of PDCR 2-050-93 PDCR 2-050-93, dated July 13,1995, was issued to install two safety-related isolating l transformers in alternate feed paths to safety-related equipment, but failed to evaluate the electrical circuit changes introduced by the transformers. Because of the addition of the new (l transformers, the circuit inpedance was substantially changed, which could have an effect on f the voltage regulation and the short circuit profiles of both redundant safety divisions of vital ac powe Safety Evaluation (SE) No. SE-2-050-93, which supported the PDCR, failed to include any objective evidence of an evaluation of the new failure modes introduced by the installation of the two safety-related isolating transformers in alternate feed paths to safety-related equipmen For example, the SE issue 3.2.1, "Effect on the probability that mitigating equipment will fail," was annotated as, "The credible failure modes are unchanged," which failed to recognize the fact that any failures associated with the new transformers would constitute new failure mode By failing to analyze any increase in the probability of failures introduced by the change, the licensee failed to thoroughly assess whether or not the change constituted an unreviewed safety question (USO).
j The failure to perform a design change engineering evaluation that included an adequate review for suitability of application of equipment that is essential to the safety-related functions of the systems and components and to include objective technical evidence to support the engineering evaluation conclusions is considered a violation of 10 CFR Part 50, Appendix B, Criterion Ill, " Design Control."
The licensee's indicated they plan to review calculation 96-01499E2 to include voltage drop calculation and evaluate the adequacy of the alternate sources for the invertors. This action is being tracked on action request (AR) No. 97010385-0 This was an ICAVP significance Level 3 finding. VIO 50-336/98-201-08, Example 2) 3.2. Review of PDCR 2-009-95 PDCR 2-009-95 failed to provide an evaluation of the impact of change of power supply to safety-related circuits for the "A" and "B" Hydrogen Analyzer power circuits. These circuits were disconnected from VA10 and VA20 buses (fed from invertors) and reconnected to VA30 and VA40 buses (fed from transformers), to obtain high short circuit current to provide for adequate coordination. However, the modification did not consider the effect of changing from an inverter-type power supply to a transformer-type power supply. The inverter-type power j supply is credited with a high reliability, constituted by the de battery source.
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Also in PDCR 2-009-95, the licensee increased the invertors' frequency tolerance bandwidth from 1 percent to 2 percent, without providing any objective evidence that indicated that the new frequency setting was tolerable and did not have any undesired effects in the operation of the connected safety-related instrumentation. By failing to analyze any increase in the probability of failures introduced by the change, the license failed to thoroughly assess whether or not the change constituted an unreviewed safety questio The failure to perform a design change engineering evaluation that included an adequate review for the suitability of the application of equipment that is essential to the safety-related functions of the systems and components and to include objective technical evidence to support the engineering evaluation conclusions is considered a violation of 10 CFR Part 50, Appendix B, Criterion ll, " Design Control."
The licensee's plan to prepare a calculation to evaluate team concerns is being tracked by AR No. 08008219-0 This was an ICAVPsignification Level 3 finding. (VIO 50-336/98-201-08, Example 3) 3.2. Review of Temporary Jumpers Jumoer Device Control Sheet No. 2/92/157 The licensee installed a bypass jumper to the alarm contacts to prevent control room nuisance alarms without attempting to determine the root cause of the group fault alarms. The reason for the jumper device was to eliminate the alarm. The alarm originated in the non-iE section of the alternate power supply to safety-related panel VA-40. The team concluded that the licensee had not implemented adequate corrective actio The licensee's failure to identify the root cause of the ground fault alarms correctly and to take appropriate corrective actions to preclude repetition is considered a violation of 10 CFR Part 50, Appendix B, Criterion XVI, " Corrective Action."
This was an ICAVP significance Level 3 findings. (VIO 50-336/98-201, Example 3) Jumoer Device Control Sheet No. 2/96/052 A temporary diesel generator (DG) was installed to provide power to safety-related loads and to allow for an extended outage of the normal EDG "B." However, the provisions for feeding the safety loads from the temporary DG did not include consideration of protection and radioactive relaying features consistent with normal operation when using the safety-related diesel generator. Since the temporary generator step-up transformer secondary winding was connected in a delta configuration, there was no source of ground fault for protective relaying to operated, which differed from the configuration grounding provided by the normal D . e
.
The failure to conduct a thorough engineering evaluation to address relevant protection requirements could have resulted in undue exposure of the safety-related equipment while connected to the temporary DG and is considered a violation of 10 CFR Part 50, Appendix B, Criterion lil, " Design Control."
This was an ICAVP signfiiance Level 3 finding. (VlO 50-336/98-201-05, Example 5) Conclusions The team identified inadequacies of low safety significance in the performance of past modifications and change processes in the areas of equipment dedication, control of documentation, nonperformance of required calculations, inadequate corrective action, and performance of safety evaluation The team found that the Parsons review check lists were adequate to address the issues of the review. The planning, conduct, and documentation of the reviews were rigorous and thorough, and appeared to have the appropriate level of detail to provide the bases for valid appraisals of the licensee's performance of change processes. The team's findings were consistent with the findings that Parsons identified in that are .0 instrumentation and Controls
. Scone of Review The team evaluated the instrumentation and controls design change processes by reviewing documents and change packages not reviewed by the Parsons ICAVP, including permanent modifications, temporary modifications, FSAR changes, drawing changes, setpoint changes, and replacement item evaluations. Other documents related to these changes, such as studies conducted as a result of EWR and vendor tests were also reviewed. The team conducted ,
walkdowns of various plant areas including the control room, plant computer room, auxiliary building, and containment building outside the biological shiel The team evaluated the following documents and change packages that were also reviewed by the Parsons ICAVP: i e PDCRs
* EWRs ;
e calculations e CDGs j e DCNs e MEPL modifications
* NCRs l * Setpoint Change Requests (SCR)
l
* specification changes Specific items reviewed are listed in Appendix l
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. Observations and Findinas 4. Independent Site Findings 4.2. Engineered Safety Feature Actuation System (ESFAS)
The licensee made significant changes to ESFAS during the last few years. Most of these changes were to resolve initi. I <bsign deficiencies such as a loss of a de vital bus causing the loss of two of the four ESFAS ensor and instrument cabinets, and a single failure vulnerability in the AFW automatic initiation system (AFAIS). Additionally, the licensee committed to additional ESFAS modifications before startup as well as during the next refueling outag Most ESFAS documents reviewed adequately addressed licensing basis issues, including a licensee-identified USQ in PDCR 2-026-93, "MP2 Engineered Safeguards Actuation System (ESAS) Modification." The USQ involved bringing two vital ac facilities together in ESAS sensor cabinets, and was identified to the NRC in letter B14788, " Proposed Revision to Technical Specifications - ESFAS," dated July 1,1994. The USQ was reviewed and resolved with a safety evaluation report (SER).
The AFAlS single failure vulnerability was identified to NRC in Licensee Event Report (LER) 94-15-00 dated June 17,1994. The licensee committed to resolve the problem by the end of the next refueling outage. The licensee replaced the analog AFAIS system with a digital system, Foxboro Spec 200 Micro System, accomplished under P-DCR 2-039-94 and approved on December 7,1994. The associated 10 CFR 50.59 SE concluded that no USQ existed, a'nd NRC was notified by letter B15070 dated December 21,1994, that the final resolution to LER > - 94-15-00 would involve a digital upgrade. The SE for PDCR 2-039-94 was thorough and took credit for extensive system verification and validation (V&V) and use of similar digital hardware at other facilitie Digital upgrades for safety-related systems was a high-visibility issue when the AFAIS upgrade design was finalized.--The staff did not issue its formal response on the issue, Generic Letter (GL) 95-02, "Use of NUMARC/EPRI Report TR-102348, ' Guideline on Licensing Digital Upgrades'in Determining the Acceptability of Performing Analog-to-Digital Replacements under 10 CFR 50.59," until after the change was issued. Under the guidance of this GL, the modification would be called a USQ. However, given that the GL was not issued, and the licensee did notify the staff of its intention to use a digital upgrade, this is a non-cited violation of 10 CFR 50.59. (NCV-50-336/98-201-12) 4.2. Electromagnetic Compatibility (EMC) The team noted that SP-M2-EE-0001, " Specification for Automatic initiation of Auxiliary Feedwater SPEC 200 Modification," Rev.1, Section 1.2, required the new AFAIS equipment to be implemented under guidance for conducted and radiated electromagnetic interference (EMI) in accordance with EPRI TRI-102323, " Guidelines for Electromagnetic Interference in Power Plants." However, Section 7C of the specification only required a certificate of conformance ;
(CoC) for testing the new equipment for susceptibility to EMI. It did not require certification to ensure that the equipment did not generate EMI at a level that would affect other equipment
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important to safety. The equipment has been installed in the plant for approximately 3 years and the licensee has not identified any EMC problems with the AFASI system. The Spec 200 Micro System has been used elsewhere by this and other licensees in safety-related applications without EMC problems. Therefore, there is little safety significance associated with not fully verifying Spec 200 Micro System EMC before installatio This lack of complete EMC testing is similar to one for the inadequate core cooling monitoring system (ICCMS) replacement performed under DCR M2-96-077. As with the AFAIS upgrade, documentation addressed the susceptibility to EMI, but nothing supported the conclusion that the new system did not generate EMI which could affect other equipment important to safet lCE-373133, Rev. O, " Evaluation of the EMI/RFI Susceptibility of the Computer Products G2VX for Millstone Unit 2," addresses the ability of parts of the new system to operate in a normal environment; it did not address the generation and conduction of EM When the team identified this, issue the licensee took prompt actions to ensure that this generated EMI/RFI will be addressed in the upcoming ICCMS validation testing. However, the new ICCMS equipment had already been connected to the vital power supply system when this problem was identified by the team on April 21,1998. Some high-speed digital equipment is j susceptible to conducted and radiated EMl; some high-speed equipment also generates EM Both the ICCMS and AFAIS are powered from safety-related power system l
The failure to account for the effects of EMI and RFI generated by the new equipment in the plant design is considered a violation of 10 CFR Part 50, Appendix B, Criterion lil, " Design i Control."
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This was an ICAVP significance Level 3 findings. (VIO 50-336/96-201-08, Example 4) The team did note that some proposed changes, such as SP-M2-EE-032, " Specification for Repair of the ESAS Level 2 Undervoltage Logic," addresses both susceptibility to and
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I emissions from electronic equipment. In other completed modifications, such as the replacement of older analog recorders with new programmable devices such as the Yokogawa uR1000 series, the licensee evaluated both susceptibility to and generation of conducted and radiate EM l l 4.2. Inadequate Human Factors Reviews The licensee failed to note that the Yokogawa uR1000 recorders installed as replacements did not have an internallight. As a result, it was necessary for operators to open the recorders and remove the fan-fold paper traces to read the recorded data. The licensee indicated the failure l to note the lack of intemal lighting resulted from performing a Human Factors Engineering (HFE) review of the recorders while they were de-energized. The human factors engineer i stated they would evaluate the lack of an internal recorder light for existing and proposed additional Yokogawa uR1000 recorders. At the time the team identified this problem, the licensee planned to use unlighted Yokogawa uR1000 recorders as qualified post-accident t instruments for TR-115, TR-125, and LR-5282 (loop 1B cold- and hot-leg, loop 2B cold- and hot-leg, and condensate storage tank level respectively) under DCR M2-97-035, " Regulatory Guide 1.97 Upgrades."
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The team identified another example of less than fully adequate HFE reviews for plant modifications. The control board labels were inconsistent with SP-EE-261, " Design Standards for Modification of Control Panels at Connecticut Yankee, Millstone Units 1,2, and 3." SP-EE- 4 261, Attachment 2, Section 1.1, " Instrument / Display Labels," details how control board labels are to be labeled. It required the use of a delimiter between the device designator (e.g., "Tl" for , temperature indicator) and instrument loop. The delimiter was a dash for all non-RG 1.97 PAM q devices, and a color coded dot for PAM instruments. Some non-PAM indicators were color i coded, and come PAM indicators had a black dash. Some indicators such as the nuclear instruments had color coded labels (i.e., "A", "B," "C," and "D") above the instruments rather i than using a dot on the label. The control room label deviations from the standard was ( indicative of a failure to perform adequate HFE reviews for change { l This was one issue addressed by DCR M2-97035 and draft DCN DM2-00-0532-98 which l changes the PAM instrument label requirement. The DCN would require PAM indications to have a triangle prefix non-PAM indicatHs would not have the triangle, and the color code would be for instruments powered from vital boards. The responsible license engineer stated the DCR would upgrade all PAM labels, but non-PAM labels which were not consistent with the revised standard and other non-PAM labels which were inconsistent with both the old and new guidelines would not be revised by the DCR. Other problems identified by the team included power range indicators Ji-005, 006, 007, and 008, and startup range indicators JKl-005, 006, 007, and 008 which were labeled in accordance with the previous specification, but are inconsistent with the new standard. As a result of this team-identified issue, the licensee stated they planned to review all control room labels and ensure that they were consistent with the new labeling guideline Procedure SP-EE-261 is a quality-related procedure. The failure to perform activities affecting quality in accordance with documented procedures appropriate to the circumstances is a violation of 10 CFR Part 50, Appendix B, Criterion V, " Instructions, Procedures, and Drawings."
This was an ICAVP significance Level 3 finding. (VIO 50-336/98-201-07) 4.2. Post-Accident Monitoring Upgrades Revising control board labels was one of the minor aspects of DCR M2-97-035. More important problems addressed by the DCR included the upgrade of many control board indicators from non-QA to safety-related status. This was done to bring the control board indicators in compliance with the commitment to RG 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environ Conditions During and Following an Accident," Rev. 2. The RG established a graded criteria for various PAM instruments; instruments classified as Category 1 had to be single failure proof and redundant or diverse indicators must be electrically independent arid physically separate. A minimum of two channels were required to meet these needs. LER 97-018-00 addressed a problem where control room indicators were not single failure proo FSAR Table 7.5-3, Rev. 49, indicated that four channels were provided for most of these
' Category 1 instruments. However, the licensee determined that only two channels were required to be upgraded and the remaining two channels would remain isolated from safety-
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related instrument loop components. The 10 CFR 50.59 SE concluded that the reduction from four channels to two channels was not a USQ, and the licensing basis was that only two channels had to be safety-related. The licensee stated that the over commitment in the FSAR resulted from a sequence of administrative errors. The team found that having only tow channels of safety-related displays was technically adequate but inconsistent with the approved licensing basis. The licensee stated they planned to address this issue in writing with the staff in the near future. After the staff reviews the submitted documents, a determination will be made if reducing the docketed correspondence of four PAM channels to two PAM channels was a USQ. This issue is identified as an Unresolve-I Item. (URI-50-336/98-201-13) 4.2. FSAR Change Process in general, the results of team reviews of 10 CFR 50.59 screening reviews and safety evaluations associated with FSARCRs showed an improving trend from inadequate to adequate as licensee standards and procedures improved over the time frame covered by the review. (This trend is similar to an observation included in the Unit 3 Tier 2/3 ICAVP Inspection Report 50-423/97-209.) However, the team identified problems with two recent 10 CFR 50.59 screening reviews. FSAR Figure 11.01-04, Sheet 1, P&lD 25203-260211, Sheet 1, " Aerated Liquid Radwaste System," was revised by Maintenance Support Engineering Evaluation l ' (MSEE) DC.N DM2-00-1102-97, " Resolution of Drawing Discrepancies for Radiation Monitoring ' Loop RM-9116 (UIR 3389)." The associated 10 CFR 50.59 Safety Evaluation Screening l Question 2, "Is the activity bounded by a previously performed 10 CFR 50.50 safety evaluation" u ! and Question 3, "Does the activity make changes to the facility as described in the SAR," were ; both answered "NO." As a result, a complete 10 CFR 50.59 SE was not performed for the
MSEE, even though the modification was associated with a change to the facility as depicted in { '
the SAR. Changes to the drawing were minor (e.g., deletion of local annunciator horns, I deletion of a handswitch, deletion of a recorder, etc), but were not drafting error I FSAR Figure 11.01-02, Sheet 1, P&lD 25203-26020, Sheet 2, " Aux Building Drains," was revised by MSEE DCN DM2-00-1104-97, " Drawing Update for Radiation Monitoring Loop RM-9049 (UIR 3352)." The associated 10 CFR 50.59 Safety Evaluation Screening Question 2 incorrectly stated that the P&lD change was contained in the scope of Engineering Evaluation M2-EV-970086, Rev. O, and was bounded by SE S2-EV-97-0204. The MSEE Safety Evaluation Screening Question 3 was also answered "NO." The MSEE changed an FSAR drawing, but was not bounded by ME-EV-70086 or its 10 CFR 50.59 S Changes to the drawing were minor (e.g., deletion of local annunciator horn and recorder), but
- were not drafting error CFR 50.59(bO(1) requires, "The licensee shall maintain records of changes in the facility and changes in procedures made pursuant to this section, to the extent that these changes in the facility as described in safety analysis report or to the extent that they constitute changes in the procedures as described in the safety rialysis report....These records must include a written
!
- safety evaluation which provides the bases for determination that the change, test, or experiment does not involve a unreviewed safety question."
l ! During the Unit 3 Tier 2/3 inspection (50-423/97-209), the staff expressed a concern with
l l
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I I l l inadequate 120 CFR 50.50 evaluations for FSAR changes. The licensee acknowledged this and stated they had taken corrective action to ensure further violations would not occur. The changes noted above occurred after the licensee had taken some corrective action to resolve the problem. The failure to conduct adequate Safety Evaluation Screening is a violation of 10 CFR 50.50(bO(1).
This was an ICAVP significance Level 3 findings. (VIO 50-336-201-06) 4.2. Environmental Qualification DCR M2-96-063, " Electrical Equipment Qualification (EEO) Modifications for SOV Circuits," issued an upgrade of seven solenoid operated valves (SOVs) located in a harsh environment to the environmentally qualified (EQ) program. This upgrade for the SOVs was appropriate. The team questioned the qualification of a sample for other solenoid valves, specifically 2-CH-506, 2-CH-516, 2-RC-001, 2-RC-45, 2-SI-621, 2-S-449, and 2-GAN-234. The licensee indicated the first five were already in the EQ program but had minor problems such as unqualified connections and the use of qualifiable cables not included in the EQ program. These problems were identified by the CMP and corrective actions were being incorporated during the inspection. Because they were identified by the licensee, occurred during an extended outage, and would be fixed before startup, the failure to properly qualify values 32-VH-506,2-CH-516, 2-RC-001,2-RC-45,2-SI-628, for an accident environment is a non-cited violatio (NCV-50-336/98-201-14) Conclusions Overall, the team found that the licensee's l&C engineers focused on safety and utilized sound engineering practices. In general, the procedures used to control Millstone Unit 2 l&C change processes coupled with industry guidance, were adequate. The team identified minor deficiencies resulting from inadequate past procedures and inadequate procedure complianc The licensee was aware of industry issues related to digital upgrades including software and EMC. However, in two cases, the licensee failed to ensure that EMI generated by new safety-related digital systems did not affect other safety-related equipmen The team concluded that Parsons had effectively implemented their Tier 3, ICAVP Audit Pla The team's findings were consistent with the findings that Parsons identified in this are .0 Structures. Pioina and Pioe Succorts Scooe of Review The team reviewed the following documents that were not reviewed by the Parson ICAVP:
* 6 DCNs l e 4 PDCRs
! * 2 DCRs e 3 EWRs
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! . ! * 2 FSAR CRs e 4 MMODs e 2 NCRs e 16 procedures ' e 10 calculations / reports e 3 miscellaneous documents The team reviewed the following documents that were also reviewed by the Parsons ICAVP: > e 2 DCNs e 2 EEWRs e 1 FSAR CR e 1MMOD e 2 NCRs e 2 procedures e 2 calculations / reports e 2 miscellaneous documents Observations and Findinas 5. Independent Site Findings 5.2. Failure to Identify Generic implementation Procedure (GIP) Outlier Conditions in Cable Trays During a field walkthrough, the team observed that several cable trays in the containment building (Z23AA10, Z22BA10, Z25AA10, and Z25AA20) were discontinuous (i.e., had gaps).
The team observed that these trays were observed to be full of cables. The gaps were located at approximately 20 percent of the typical span between two supports. Therefore, a cantilever condition existed at the adjacent ends that was approximately 80 percent of a span. There was noticable deflection at the ga The site document that governs the seismic qualification of cable tray systems in the GIP for Seismic Verification of Nuclear Plant Equipment," Rev.2, dated February 14,1992. The GIP provides the detailed technical approach generic procedures and documentation used to provide resolution of GL 87-02/USl A-46. The following were identified as being in conflict with the GIP procedur (1) The longer cantilevered portion of the cable tray exceeded in the one-half span or i approximately 5 feet (GIP 8.2.2 Rule 1).
(2) Sag of cable tray appeared to be greater than 1-inch and did not meet the intent of GIP 8.2.3, "Other Seismic Concems."
(3) GIP 8.2.3, Concem 5, on missing components may not have been met.
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The subject cable trays should have been identified as a GlP outlier (since it did not meet the f screening guidelines of the GlP) and addressed accordingly to meet the requirements of Section 5.0 of the GI CFR Part 50, Appendix B, Criterion ill, requires, in part, that design control measures be provided for verifying or checking the adequacy of design. Failure to correctly identify the cable tray gaps as an outlier to the GIP and address them as required by Section 5.0 of the GIP is a violation of 10 CFR Part 50, Appendix B, Criterion Il In response to this finding, the licensee generated CR M2-98-1099. The licensee stated that the corrective action willinclude an evaluation of the cantilever condition of the cable tray bank for structural adequacy as well as a review of the USl A-46 walkdown informatio This was an ICAVP significance Level 3 findings. (VIO 50-336/98-201-05, Example 3) 5.2. Failure to Properly Document independent Verification of Critical Portions of a Calculation Pipe suppen calculation M2505194-1649-C2, Rev. O, which was performed to support MMOD M2-97508 was reviewed to confirm that the calcu!ation process was properly implemente The calculation has 582 pages of which 576 are Attachments A through I. Several of the attachments were either not signed by the preparer or the reviewer or the reviewer signature was provided by someone other than the reviewed. Specifically, Attachment G, which provides important design inputs / calculations that were used to qualify the support was not signed by either the preparer or checker. Section 2.0, " Conduct of Independent Review," of the Design Control Manual Chapter 4, requires the independent reviewer to complete and sign from A-1 A, an independent reviewer checklist. Section 4.0, "New Calculation Preparation / Approval Task 14," specifically states that form 4-1 A is not required to be attached to the calculation and apparently is not required to be saved. Therefore, there is no record of the form ever being complete NNECO responded that Chapter 5 of the Design Control Manual requires the verification of design inputs in accordance with Chapter 4, " Design Verification," however, a completed form 4-1 A is not required for calculation inputs. Objective evidence of design input verification is accomplished by the independent review sign off on the calculation cover sheet. The licensee c. ted that the observed inconsistency (i.e., attachments to the calculation was signed by the performer and the checker) is acceptable with the objective evidence requirements of calculation design input verification as described above. However, the licensee stated that the verification process for calculations with multiple parts will be reviewed by the DCM working group and is being tracked by AR 98008242-01 to determine if this process should be strengthene The team feels that there is nothing in the existing Design Control Procedure that allows the use of results from unchecked calculations. Furthermore, the team disagrees with NNECO's position that the use of unchecked calculations is acceptable. As a minimum, it appears that there is a weakness in the Design Control Process that does not require objective evidence that an important design input has received proper documentation and that the calculation was
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properly performed and verified / checked. The NNECO response raises a greater concern that they may not perceive that unchecked or undocumented review is a concer The failure to provide evidence that the calculation was properly reviewed and that design control measures provided for verifying or checking the adequacy of design, is a violation of 10 CFR part 50, Appendix B, Criterion ill, " Design Control."
This was an ICAVP significance Level 3 finding. (VIO 50-336/98-201-5, Example 4) 5.2. Condensate Storage Tank Design Code of Record The original condensate storage tank (CST) was fabricated as an atmospheric tank in accordance with ANSI /AWWA D100, "American Water Works Association Standard for Welded Steel Tanks for Water Storage." The CST was then changed to a pressurized tank with a nitrogen blanketing system in addition to other modifications detailed in PDCR 2-079-92. The PDCR states that the CST will be provided with adequate pressure-relief devices and a sufficient nitrogen supply to control the nitrogen blanket within a pressure range of +1 psig to 0.34 psig. Calculation 90-032-423-EC(2) uses provisions from API Standard 620,
" Recommended Rules for the Design and Construction of Large Welded Low-Pressure Storage Tanks," and API Standard 650,." Welded Steel Tanks for Oil Storage," to qualify the tank for the intemal pressure of +1.0 psi. The API Codes limit the extemal pressure to 0.0625 psi (API-620 Para 3.10.5.3). API-620 and API-650 Codes do not provide specific guidance for external pressure loadings. - They also do not provide guidance on the design and analysis for intermediate stiffening rings except for rings required to resist wind and seismic horizontal loads. For the evaluation of external pressure loadings, the calculation used ASME Ill, Subsection NE, which provides guidance for external pressure loadings on structures similar to the CS '
FSARCR 97-MP2-97 ltem 2 proposes change to Tables 1.2-1 and 4.2-4 to address the CST classification change (pressurized versus atmospheric) and changes to the applicable design codes (API versus AWWA). However, since the API Codes do not provide guidance for l extemal pressure, the ASME Code should also have been identifie CFR Part 50, Appendix B, Criterion lil, " Design Control," requires, in part, that, " Measures 1 shall be established to assure that applicable regulatory requirements and the design % ,g - basis...are correctly translated into specifications, drawings, procedures, and instructions."
<
; it appears that the licensee failed to translate correctly the design-basis requirements for the iCST design Code contained in FSAR Tables 1.2-1 and 4.2-4 into the design document ~ Calculations that qualified the CST is a non-cited violation. (NCV-50-336/98-201-15)
in response to this finding, the licensee generated ECR 25203-ER-97-0265, Rev.1, and j FSARCR 97-MP2-97 was revised to include the proper ASME reference d l
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. Conclusions The teare, concluded that in the area of structural piping and pipe supports, Millstone Unit 2, has maintained their design and licensing bases.- Areas of weakness were identified in the implementation of the GlP program and in the adherence to the proper implementation of the design control system for calculations, however, these findings were not safety significan The team concluded the Parsons has effectively implemented their Tier 3, ICAVP Audit Plan and Procedure The team's findings agreed with the findings that Parsons identified in this are .0 Ooerations The team evaluated changes to operations, maintenance, and surveillance procedures to determine if the changes maintained the Unit 2 design and licensing base An in-depth review of the Plant Operational Review Committee (PORC) approved abnormal operating procedure (AOP) 2582, Rev.1, " Loss of Spent Fuel Pool Cooling," and compared AOP 2582, Rev.1, to engineering design drawings, licensing bases, and calculations. The team determined that the procedure was consistent with the existing plant configuration and design base The team determined that EOP 2537, " Loss of All Feedwater," was concise and easily readable. EOP 2537 was a symptom-based procedure developed by following the generic Combustion Engineering Owners Group (CEOG) CEN-152 procedure guidance. The team found that the licensee appropriately modified the CEOG generic procedure to agree with the configuration of Millstone Unit 2. A step deviation document was prepared and describes in detail why and where a deviation from the CEOG-approved procedure exists. In general, the deviations were a result of inserting the plant-specific information for components. Samples of specific instruments setpoint changes were reviewed and determined to be consistent with the design bases and appropriately incorporated into this Millstone Unit 2 EO The team determined by comparing the past three executed surveillance procedures with the Technical Specifications that SP-2602A, Rev. 4, " Reactor Coolant Leakage," was adequat Overall, the team concluded that the reviewed changes did not place Millstone Unit 2 outside of its licensing and design base .0 MEPL Safety Classification Chanaes The team advised Millstone Unit 2 management that the MEPL, also known as the "Q" list, needs additional attention. The facility presented a sample plan that used Mil Std.105 to scope the problem of inappropriate parts installed in the plant. The basis of the sample plan, Mil St was not a proper sample technique. Mil Std.105 is typically used for common parts from a common lot and if a failure is determined when the acceptance criteria accepts O defects, then the entire community is 100 percent tested. The facility plan sampled a mix of components and
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did not have an adequate plan to expand the sample size when failure is detected. The facility sample of 87 installed items required replacement of two component The team had a subsequent meeting regarding the status of the MEPL with the licensee and concluded that MEPL could have a significant impact on the recovery of Unit 2. The licensee committed to develop an action plan and submit this plan to NRC. This issue is being closely monitored by the Millstone Unit 2 NRC Senior Resident inspector and is identified as an Inspector Followup Item. (IFl 50-336/98-201-16) Entrance and Exit Meetinas The team conducted an entrance meeting on April 13,1998, for the Millstone Unit 2 Tier 3 inspection. On May 4,1998, the team conducted an entrance meeting at the Parsons offices in Reading, Pennsylvania. During each of these meetings, the team discussed the scope, duration, and expected support requirements for each phase of the inspectio On June 12,1998, the team leader conducted an exit meeting at the Millstone Training Facility which was open for public observation. During this meeting, the team's findings and observations were discusse A partial list who attended these meetings is attached in Appendix B.
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Appendix A List of Apparent Violations, Unresolved items, and Inspector Followup Items This report categorizes the inspection findings as violations (VIO), apparent violations being considered for escalated enforcement (EEI), unresolved items (URis) or inspector followup items (IFI) in accordance with Chapter 610 of the NRC Inspection Manual. An apparent violation is a matter about which the Commission has concluded there is enough information to conclude e violation of a legally binding requirement has occurred. The violation is classified as apparent until the NRC assigns a severity level and the licensee is given the appropriate chance to respond to the NRC's determinations. A URIis a matter about which the Commission requires more information to determine whether the issue in question is acceptable or constitutes a deviation, nonconformance, or violation. The NRC may issue enforcement action resulting from its review of the identified URIs. An IFl is a matter for which additional information is needed that was not available during the inspectio Item Number Finding Section(s) Title Type 50-336/98-201-01 VIO 2.2. Failure to perform leakage testing of safety-2.2. related valves in systems that could contain highly radioactive fluids during an accident in accordance with 10 CFR Part 50, Appendix B, Criterion X /98-201-02 VIO 2.2. Failure to implement adequate corrective 2.2. actions in accordance with 10 CFR Part 50, 3.2. Appendix B, Criterion XV /98-201-03 VIO 2.2. Design requirements of ASME Section Vlli, Article UG-134(a), for pressure-relief devices, were not correctly translated into the design for the RBCCW heat exchangers' relief valves' setpoint j 50-336/98-201-04 VIO 2.2. Failure to implement design control measures for design drawings in accordance with 10 CFR Part 50, Appendix B, Criterion Il I 50-336/98-201-05 VIO 3.2. Failure to implement design control measures 3.2. for design changes in accordance with 10 CFR l
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, 5.2. Part 50, Appendix B, Criterion Il ' l 5.2. /98-201-06 VIO 4.2. Failure to perform adequate safety evaluations ! in accordance with10 CFR 50.59, " Changes, I tests and experiments."
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50-336/98-201-07 VIO 4.2. SP-EE-261was not followed for changes made to control room panel labels contrary to 10 CFR Part 50, Appendix B, Criterion /98-201-08 VIO 3.2. Failure to implement design control measures 3.2. without ensuring the suitability of the new 3.2. equipment for its intended use in accordance 4.2. with 10 CFR Part 50, Appendix B, Criterion Il /98-201-09 IFl 2.2. Review RWST backleakage through various flow paths such as 2 CS-28 as described in Memo TS2-98-0122, dated April 9,199 /98-201-10 IFl 2.2. Review of ECCS pumps (HPSI P-41 A, B, & C and LPCl P-42A & B) seal o-ring material qualification requirements for radiation and temperatur /98-201-11 IFl 2.2. Capability of the enclosures for MCCs B51 and B61 to withstand the overpressure from the most severe HEL /98-201-12 NCV 4.2. Failure to classify analog-to-digital replacement modifications under 10 CFR 50.59 as a USO
.50-336/98-201-13 URI 4.2. Determination will be made if reducing the docketed correspondence of 4 PAM channels to 2 PAM channels was a USQ. The team found that having only two channels of safety related displays was technically adequate but inconsistent with the approved licensing basi /98-201-14 NCV 4.2. Failure to properly qualify solenoid valves 2-CH-506, 2-CH-516, 2-RC-001, 2-RC-45, 2-SI-628 for an accident environment. Verification of appropriate operator actions following a auxiliary feedwater system pipe break upstream of the cavitating ventur /98-201-15 NCV 5.1.1.3 Failure to correctly transtate the design basis requirements for the CST design code contained in FSAR Table 1.2-1 and 4.2-4 into the design documents and calculation /98-201-16 IFl The licensee committed to develop an action plan regarding the status of the Millstone Unit 2 MEPL and submit this plan to NRC. This issue is being closely monitored by the Unit 2 NRC Senior Resident inspecto A-2
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Appendix B Entrance and Exit Meeting Attendees (Partial List)
NAME ORGANIZATION Northeast Nuclear Enerav Comoany M. Bowling Unit 2 Recovery Officer J. McElwain Unit 2 Recovery Officer P. Loftus . Manager, Regulatory Affairs M. Ahem Manager, Design Engineering D. Harris ~ _ Coordinator, Regulatory Compliance J. Fougere Manager, ICAVP R. Necci Director, Configuration Management Program R. Laudenat ICAVP Program Director, Regulatory Affairs S. Brinkman Director, Unit 2 Engineering . R. Boehling - Asst. Director, Unit 2 Engineering
F. Mattioli Supervisor, ICAVP J. Price Director, Unit 2 R. Ewing Supervisor, Design Engineering, Unit 2 K. Fox Supervisor, Engineering, Unit 2 R. Joshi . Manager, Regulatory Compliance, Unit 2 G. Komosk ICAVP inspection Lead, Design Engineering B. Wilkens j l ! Manager, Programs and Engineering Standards R. Lawrence Representative, ICAVP R. Bonner Engineuing Supervisor, Unit 2 Operations r J. Pizzi - Representative, ICAVP l i R. Crittenden Representative, ICAVP M. Bain - Manager, Technical Support Engineering M. Flasch Manager, Recovery Oversight i P. DiBeneregio : . Director A &P, Nuclear Oversight M. Heal Lead, Nuclear Oversight Connecticut Nuclear Enerav Advisorv Council J. Markowicz Representative _ US Nuclear Reaulatorv Commission E. Imbro . NRC/ Deputy Director, ICAVP, SPO 1 L S. Reynolds NRC/ Chief, ICAVP,SPO I l V. Ferranini NRC/ Contractor- Team Member j D. Prevatte NRC/ Contractor- Team Member " 8.Hughes NRC/lCAVP, SPO - Team Member O.Mazzoni NRC Contractor- Team Member R. Quirk NRC Contractor-Team Member H. Eicherholz NRC/ICAVP, SPO
' D. Beaulieu NRC Senior Resident inspector - Unit 2 l
S. Jones Resident inspector - Unit 2
' R. McIntyre System Lead, Team 28, SPO l
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Appendix C List of Documents CALCULATIONS AND REPORTS e Calculation 96-ENG-1502M2 " Seismic Bracing of DC Switchgear Instrument Panels VR11, VR21, , D11, D12, D21, D22 to resolve USl A-46 (G.L 87002) & IPEEE (G.L. 88-02, Supp. 4) Outlier," Re * Report No. 90008-01 " Seismic Analysis of a Masoneilan Model 35-35112 Camflex Control Valve," December 7,199 * Report No. PA 96984, " Seismic Qualification by Analysis G.H. Bettis Model t316-SR2(CW)- , M3HW-10S Operator /14," 150 lb PermaSeat Valve" Rev. e Calculation M2406102-01632-C2, " Calculation for RWST Pipe Support 406102," Rev. * Calculation M2401032-01599-C2, " Calculation for RWST Pipe Support 401032," Rev. O e Calculation M2505194-01649-C2, " Calculation for RWST Pipe Support 25203-22200-505194," Rev. * Calculation M2505194-01649-C2, " Calculation for RWST Pipe Support 25203-22200-505194," Rev. * Calculation 90-032-422-EC(2), "MP2 CST N2 Blanket - Seismic Review Plus Modifications Part 1," Rev. O, including CCN 1 & * Calculation 90-032-423-EC(2), "MP2 CST N2 Blanket - Seismic Review Plus Modifications Part 2," Rev. O, including CCN * Calculation 8011-5.9, 'USNRC l.E. Bulletin 80-11 Design Verification of Block wall 5.9," Rev. * Calculation 89-078-01722M2," Millstone Unit 2 Minimum Required Stem Thrust Calculation for MOV 2-SI-651 and 2-SI-652 Using EPRI PPM Methodology," Rev.0, including CCN 02 and 0 * Calculation Vectra MP2OR, Section 5.9, " Anchorage Detailed Evaluation M2 Tank T48A & TA488," Rev. * 90-032-423-EC(2), MP2 CST N2Blanket - Pressure Review Plus Modifications - Part 2,Rev. O, May 18,199 * 89-078, Millstone Unit 2 Pressure Locking and Thermal Binding Evaluatior, of Gate Valves, Rev. O, June 27,199 * M2-EV-960031, Evaluation of Diesel Oil Day Tank T-48B Low Level Alarm Problems, Re * M2-EV-97-013, initiation of Auxiliary Feedwater to Affected Steam Generator Following Main Steam Line Break AccidentM2-EV-970086, Rev.0, Technical Evaluation for Non-Design Basis Changes to P&lDs that are incorporated in the FSAR as Figures, Re * MSEE-DM2-00-1102-97, Resolution of Drawing Deficiency for Radiation Monitoring Loop RM-9116 (UlR 3389).
- MSEE-DM2-00-1104-97, Resolution of Drawing Deficiency for Radiation Monitoring L oop RM-9049(UIR 3352).
- MSEE-DM2-00-1203-96, Existing ESAS Actuation Module Drawin * 18767-ICE-373133, Evaluation of the EMI/RFI Susceptibility of the Computer Products G2VX for Millstone Unit 2, Rev. )
, DCN LIST
* DCN DM2-S-0632-96, "USI A-46 Seismic Interaction for Control Room Cabinets," Rev. * DCN, "USI A-46 Seismic Interaction Resolution for Panels D11, D12, D21, D22, VR11, VR21,"
* DCN DM2-07-0217-97, " Restraining of 10" HBD-57 Blow Out Piping-MSSV's SG. No 1" C-1 I
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e DCN DM2-00-1043-97, "HPSI Vent Line (2-SI-107 & 41 A) Modification in Support of 2-RB-37.2B MOV Actuator", DCN DM2-S-0103-96, " Valve 2-LRR-8 Remote Operator"
* DCN DM2-00-1274-96, " Concrete Wall Repair in Diesel Generator Room 8," October 3,1996. *
e DCN DM2-01-0128-97, " Intake Structure Chkd. Plate and insert Frame Detail," March 1,1997. *
* DCN DM2-00-1369-97, "HPSI Discharge Pipe Support Drawing Update NCR 297-530,"
December 9,199 * DCN DM2-00-1466-096, Drawing 25203-30024, " Lack of controlled document for breaker instantaneous settings," Deember ?1,1996. *
* DCN DM2-00-0562-97, " Diesel Day Tank Low Level Alarm Setpoint Change."
e DCN DM2-S-0010-96, "ESAS Logic Diagram Update."
DCR - PDCR LIST e DCR M2-97020, " Steam Generator Snubber Control Valve Upgrade" Rev., * DCR M2-97056, " Stem Replacement for 2-SI-652," Rev. * DCR M2-96-063,"EEQ MODIFICATIONS FOR SOV CIRCUlTS."
e DCR M2-96-065, " Pressurizer Level QA Indication."
e DCR M2-96-558,"RCP 408 Seal Pressure HI Alarms Setpoints Retum to Normal."
e DCR M2-97-021, " Diesel Oil Supply Tanks Low Level Alarm Setpoint Change."
- DCR M2-97035, " Regulatory 1.97 Upgrades."
- DCR M2-97057, " Millstone Meteorological Tower Temperature Instrument Replacement."
e PDCR 2-79-92, " Condensate Storage Tank (CST) Nitrogen Blanket." Re e PDCR 2-031-93, " Condensate Storage Tank Pipe Trench Support 405286 Modification," Rev. * PDCR 2-175-92, "'A' Charging Pump Motor Anchorage Modification," Rev. e PDCR 2-205-92, " Support of Actuators for Control Valves 2-ES-79 C, E, G, & 1," Re * PDCR 2-007-88, "Auxilairy Feedwater System Steam Bypass Line," Rev.1.*
* PDCR 2-114-92, " Required Modifications for Main Steam Line Break Scenario," Rev. O. * * PDCR 2-015-92, " Modify the MP2 Service Water Pump Bearings, Shaft Sleeves, Wearing Rings, and Lube Water Systems," Rev. * PDCR 2-79-92, " Condensate Storage Tank Nitrogen Blanket," Rev. e PDCR 2-136-92, " Modification of Shutdown Cooling Suction Valve 20SI-652 to Prevent the Possibility of Hydraulic Lock," Rev. * PDCR 2-160-92, " Auxiliary Feedwater Pump Improvements," Rev. e PDCR 2-147-92, "HPSI Pump Mechanical Seal O-Ring Material Change," Rev. * PDCR 2-013-93, " Modification to Vital DC Switch Gear Room Temperature Controllers and Temperature Control Valves," TIC-8864, TV-8864 (2-CHW-34), TIC-8867, & TV-8867 (2-CHW-4), Rev. * PDCR 2-208-92, " Coat Service Water Pump P5C Columns with Belzona," Rev. * PDCR 2-168-92, " Modify Service Water Valve 2-SW-13A & 2-SW-138 Check Valve Bushings,"
Re e PDCR 2-171-92, " Atmospheric Dump Valve Gasket Modifications," Rev. * PDCR 2-047-95, " increase 2-SW-89A and 2-SW-89B Opening Stroke Time," Rev. * PDCR 2-034-95, " Replacement of 2-SW-111 and 2-SW-113," Rev. O.
l e DCR M2-97007," Vital Coolers for MCC B51 and B61 Enclosures," Rev. O, November 22,199 e PDCR 2-064-95, " Air Accumulator for 2-RB-13.1 A & B," Rev.1, August 20,199 * PDCR 2-052 95, " Plugging Drilled Hole c in HPCI Pump Suction Valves' Disks," July 23,1995. *
* PDCR 2-023-95, " Drilling Holes in HPCI Pump Suction Valves' Disks," March 13,1995.*
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% . * PDCR 2-7-90, " Install Thermal Overload Relays in all Motor-Operated Valve Starters in Safety-
! Related MCCs," May 14,1990.
{ * PDCR 2-023-94, " Reconnect Zenner Diodes Z1 and Z2 in DC open pole detection scheme to ! restore operability of the circuit," April 29,1994.
i ' e PDCR 2-50-93,7/13/95, Changes to 120 VAC vital system e PDCR 2-009-95,7/18/95,120 VAC vital system electricalisolation changes e DCR M2-96057,6/20/96, Damper Motor Replacement
* PDCR 2-001-92,2/11/92, SBO AAC Power Supplies for MP1 and MP2 * * PDCR 2-006-92, Replace Under-rated ASCO SOVs e PDCR 2-010-94, Pzr Pressure Control SPEC 200 Wiring Modification l * PDCR 2-012-94, RCP D Seals Pressure Setpoint Changes
' e PDCR 2-013-94, RCS Cold Leg Wiring Changes l e PDCR 2-016-94, RCP D Upper Lube Oil Reservoir Setpoint Change l t l * PDCR 2-017-93, AFW Flow Indication e PDCR 2-026-93, ESAS Modification - Mostly Power Supplies
* PDCR 2-029-94, RCP Upper Lower Lube Oil Res. Setpoint Change e PDCR 2-037-94, MP2 Control Room Annunciator General Silence Switch e PDCR 2-038-95, RCS Flow Instrument Modification * PDCR 2-039-94, AFW Auto Initiation Control Mod
" e PDCR 2-071-92, Diesel Control System Modification
* PDCR 2-108-92, Diesel Oil Storage Tank T47A Low Level Setpoint I e PDCR 2-139-92, Vent High Range Rad Monitor Setpoint Change e PDCR 2-182-92, Pressurizer Pressure Setpoint Vnitage to Current Converter Modification e PDCR 2-45-82, TM/LP Pre-Trip Setpoint Change o PDCR 2-95-76, ESF Pumps IVJ,imum Flow Bypass EWR LIST e EWR 96-0112, " Plant Equipment Modifications in Response to GL 87-02/USl A-46 " Rev. * EWR # 2-95-054, " Radiation Monitor # 8434 Supply Tubing From Aux. Build;ng Ventilation Near 2-HV-123," Decemer 1,1996. * * EWR # 2-95-097, " Auxiliary Building Roof Penthouses - Installation of Snow Screens / Barriers,"
September 18,1995.*
* EWR 2-92-A0172, "RBCCW Overpressure Relief, February 1,1996. * * EWR M295316, "Lonegran Relief Valves," March 1,1996. * * EWR M2-97139," Service Water Discharge Strainer Automatic Backwash High DP Alarm Setpoint Change," November 20,1997. * * EWR 96-0153, "RCP "C" Pump Seal Replacement," October 24,1996. *
e EWR M2-96078, " Loss of RBCCW Inventory Through Backflow, April 8,1996. *
* EWR M2-96-136, " Single MSIV Closure Time," April 25,1997. * * EWR M296076, " Replace 12 Primary System Thrmowells," June 22,199 * M2-97182, " Replace overloaded cables in trays," September 22,199 * EWR 2-95-012, " Replacement of ICCMS."
- EWR 2-95-016, "Pzr Level RG 1.97 Indication Mod for QA Power."
- EWR 2-96-142, " Start-up Rate Trip."
- EWR 2-97-008," Revise Pressure Switch Setpoint."
e EWR 2-97-065, " Diesel Day Tank Low Level Alarm Setpoint."
l e EWR 2-97-099, "RG 1.97 Loop Upgrade."
e M2-97116,2/27/98, " Interaction between ATI load Sequencer test and RSST."
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e 2-94-00059, " Update cable ampacities," December 31,199 * 2-93-00072, " Splice inverter cable," June 7,199 FSAR CHANGES l
* FSARCR 97-MP2-97, " Text and Table Changes Related to the Auxilary Feedwater System."
e FSARCR 92-MP2-20, " Tendon Surveillance Change to Comply with Regulatory Guide 1.35," January 20,199 * FSARCR 88-MP2-24, " Modify Table 5.2-12 To indicate Spare Containment Penetration #34 Was Used to Provide Backup Instrument Air To Safety Related Valves," January 20,199 * 95-MP2-22,8/12/95, "FSAR Section 6.3.3, System Operation (provides explanation for maintaining containment sump piping full of borated water)," August 12,1995. *
* 85-MP2-13, "FSAR Section 14.20.3.2, Control Room Habitability," May 4,1987. * * 94-MP2-7, " Spent Fuel Storage," April 21,1994. * * FSARCR 92-MP2-34, "ESFAS Instrumentation" * FSARCR 93-MP2-14, "ESFAS."
- FSARCR 95-MP2-26,"ESFAS SRAS Start /Auctioneered Power Supplies."
- FSARCR 95-MP2-33, "ESAS/AFAIS."
- FSARCR 97-MP2-100,"RCS Design Bases and System Description."
- FSARCR 97-MP2-109, " Pressurizer Level and Bottle-up Panels."
- FSARCR 97-MP2-34, "ESFAS - ventilation."
- FSARCR 97-MP2-41, " Containment Pressure."
- FSARCR 97-MP2-56, "ESFAS System Description and SRAS."
MMODS i e MMOD M2-96569, " Modification of Control Room Cabinets to Prevent Seismic Interaction," Rev. e MMOD M2-96570, " Modification of DC Switchgear Rooms Distribution Panels VR11, VR21, D11, D12 D21, and D22," Rev. e MMOD M2-97508, " Axial restraint at RWST Pipe Support #505194," Rev. e MMOD M2-97508, " Axial Restraint at RWST Pipe Support #505194," Rev. * MMOD M2-97-510, " Snubber to Strut Replacement for 4 RWST Pipe Supports," Rev. NCRs
* NCR M2-96-0082, June 1,199 * NCR M2-96-243, " Failed Hilti Bolt on RWST Train A Support # 401032."
e NCR-2-96-0197, " Pipe Support #301080 Discrepancies on the RWST Supply System Identified as a Result of Engineering and ISI inspections (Cracks in Concrete Extends from Baseplate to Baseplate of Support # 401004," December 20,1996. *
* NCR 296-248 " Eroded Valve Seat on 2-SW-241, " Continuous Service Water Min-Flow to Diesel isolation Valve," November 13,1996. * * NCR 294-158, " Vital Switchgear Emergency System (Various Components)," June 2,199 * NCR 295-379, " Improper Downgrade of F39A, FV-8123, & FV-8262," June 27,199 * NCR 292-1074, " Millstone 2 Service Water," May 13,199 * NCR 2-84-0064, "2-MS-365 Has A Packing Leak," June 16,1985.* * NCR 297-080, "2-SI-114 Has Several Gouges On the Underside of the Valve Cover," October 21, 199 C-4
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o NCR 2-91-0225, " Safety injection Tanks High & Low Pressure Abnn Switches," September 30, 1991.*
* NCR 294-244, " Motor Control Center Breaker B6168," Nevamber 17,199 * NCR292-276," Battery Charger DC Bkrs do not meet acceptance criteria of PT 21421D," August 4,
' 199 * NCR293-011, " Replacement of a QA bushing with a Non-QA bushing," January 9,199 l
* NCR2-154-79," Cable upgrade to QA," October 15,197 * NCR-295-308, " Temporary Motor Heaters for B-LPSI Motor," May 5,1995. *
STATION PROCEDURE OR FORM CHANGE e MP2721Z, " Application of Belzona Molecular Materials," Rev. O, Change 3, June 20,1989. *
* SP 2605D, " Containment Leak Test, Type C," Rev.15. * * SP-EE-261, " Design Standards for Modificiiton of Control Panels at Connecticut Yankee, ,
Millstone Units 1,2, 3." l
* SP-M2-EE-012, " Design Specification for RG 1.97 Instrumentation."
PROCEDURES
* C EN 101c, " Management of ASME Section XI Inservice Inspection Programs," Rev. O., March 1, 199 * WC 3 "ASME Section XI, " Repair and Replacement Program," Rev. O., Change No. e 11867-014-P002, " Procedure for Pipe Support / Restraint Design Review by Contractor," Rev.1, November 29,197 * 11867-014 P003, " Procedure for Piping Stress Analysis by Subcontractor," Rev. 2.,
December 18,197 * " Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment," Rev. 2., February 14,199 * SQ Review Checklist SQR 2-96-019, " Modification of Control Room Cabinets to Prevent Seismic Interaction and Achieve G.L. 87-02/USl A-46 Outlier Resolution."
e " Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment," Re * Specification SP-ME-730," Specification for Qualification of Small Bore Piping and Tubing Systems for Deadweight, Thermal, and Seismic Loads," Rev. * 25203-SP-CE-132 R2, " Technical Specification for Turbine Bldg.. Structural Modifications," Rev. 2., May 1,1984. *
* IST Procedure SP2611A, Rev. 7, and SP2604B Rev 7, "RBCCW Pump Operability and inservice Testing Facility 1 and Facility 2 (Pump B)". * * ISI Procedure NU-UT-1, " Ultrasonic Examination, General Requirements," Rev.13. * * ISI Procedure NU-UT-2, " Ultrasonic Examination of Piping Welds," Rev.12, and 2 change * NGP 2.32, " Nuclear Group Procedure Engineer Programs," Rev. e NPG 5.19, " Nuclear Group Procedure Seismic Qualification Reviews," Rev. * NPG 5.29, " Nuclear Group Procedure The Design Control Process," Rev. * NPG 6.12, " Nuclear Group Procedure Evaluation Of A Replacement Item," Rev. * NPG 2.29, " Nuclear Group Procedure Justification For Continued Operation," Rev. * NPG 2.01," Nuclear Group Procedure implementation Of 10CCF21 Reporting Of Defects and Noncompliance," Rev. 9.
- e OP2315D, " Vital Electrical Switchgear Room Cooling Systems, Rev.10, April 17,1998.
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* SP 2606D,"Containtnent Spray System Alignment, Facility 2," Rev.1 e EOP 2532, " Loss of Primary Coolant," Rev.15.
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o MP 27212, " Application of Belzona Materials," Rev. e Millstone Nuclear Poswer Station Design Control Manual (DCM) Vol. * CMP 762A, " Control of Temporary Electrical Power."
e WC10, " Jumper, Lifted Lead, and Bypass Control."
e - WC5, " Fuse Control."
e NGP3.12, "10CFR 50.59 Process."
- NGP4.02, " Proposed Technical Specification Change Requests and Requests for Enforcement Discretion."
e NUC MPM 3.01, " Commercial Grade Dedication."
- NUC MPM 3.02," item Equivalency Evaluation."
- e RAC-02, " Technical Specification Change Requests and implementation of License Amendments."
e RAC-12, " Safety Evaluation Screens and Safety Evaluations."
DRAWlNGS e P&lD 26022, RBCCW System, RBCCW Pmp and Heat Exchangers, Re * P&lD 26002, Sheet 1, Main Steam From Generators, Rev. 40, January 28,199 e P&lD 26002, Sheet 3, Main Steam Turbine & Reheater No.1 A, Rev. 23, December 8,199 * P&lD 26002, Sheet 4, Main Steam Turbine, Rev.17, July 8,199 > e P&lD 26005, Sheet 3, Condensate Storage and Auxiliary Feed, Rev 31, November 9,199 e P&lD 26008, Sheet 2, Service Water, Rev. 60, November 8,199 e P&lD 26015, Sheet 1, Low pressure Safety injection System, Rev.19, November 10,199 * P&lD 26015, Sheet 2, High Pressure Safety injection Pumps, Rev.15, January 29,199 * P&lD 26015, Sheet 3, Safety injection Tanks, Rev.13, August 1,199 * P&lD 26022, Sheet 1, R.B.C.C.W. Pmp and Heat Exchangers," Rev. 34, January 29,199 * P&lD 26027, Sheet 2, Turbine Budg., intake Str., Whse. & D.G. Rms. Chilled Water System, Revc 32, October 30,199 * P&lD 26033, N 2System, Rev. 27, February 5,199 e P&lD 26029, Sheet 1, Auxiliary Building Ventilation System, Rev.15, November 5,199 e PalD 26008, Sheet 3, Service Water To Vital AC Switchgear Cooling Coil and AC Chillers, Rev 18, November 8,199 e 25203-30006 Sheet 1, Single Line Metering and Relay Diagram 4160V Emergency Diesel Generator H7A.
- e 25203-30006 Sheet 2, Single Line Metering and Relay Diagram 4160V Emergency Diesel
. Generator H7 e ~25203-30002, Single Line Metering and Relay Diagram Generator No. 2 Main, Norm and Reserve Station Service Transformers e 25203-30005, Single Line Meterir:g and Relay Diagram 4160V Emergency Buses 24C,24D e 25203-30004, Single Line Metering and Relay Diagram 4160V Emergency Buses 24A,24B e 25203-30004, Single Line Metering and Relay Diagram 4160V Emergency Buses 24E,24G e 25203-30003, Single Line Metering and Relay Diagram 6.9 kV Buses 25A,258 e _25203-30007, Single Line Diagram 480 V Unit Subst 22A,228,22C,22D e 25203-30008, Single Line Diagram 480 V Unit Subst Emerg 22E,22F e 25203-30023, Single Line Diagram 125VDC System o 25203-30024, Single Line Diagram 125VDC Emergency and 120VAC Vital Syst e 25203-30001, Main Single Line Diagram e 25203-26015 Sheet 1, P&lD - Low pressure Safety injection C-6
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TECHNICAL SPECIFICATION CLARIFICATION AND CHANGE RREQUESTS e TRMCR 95-2-17,3/4.3.2 ESFAS Instrumentation, July 10,199 e TRMCR 95-2-14,3.7.1.2, Auxiliary Feedwater Pumps, June 23,199 * TRMCR 95-2-9, 3.6.1.2, Containment Leakage, June 16,199 * TRMCR 95-2-18, 3/4.6.1, Primary Containment, July 11,199 e PTSCR 2-13-97,3/4.8, Bases - Electrical Power System * PTSCR 2-22-93, ch Spec 2.2.1 Table 2.2-1 Correction e SCR M2-05-015, P B Mid and Upper Seal Pressure High Alarm , e SCR M2-95-020,1 SG Low Level e SCR M2-96-001, Jacket Coolant Temperature Controller Setpoint l e SCR M2-96-003, Jacket Cooling Low Alarm Setpoint !
* SCR M2-96-006, Control Room AC Suction Pressure Cut Out e TRMCR 95-2-19,12/22/94, Add requirement to surveillance 4.8.1.1.1 to verify offsite source breaker alignment e PTSCR 2-13-95,05/23/95, only one channel of ac. and dc. power is required to be verified operable for modes 5 and 6 * PTSCR 2-2-95,03/15/95, EDG load and sequencer surveillance not required for modes 5 and 6 * PTSCR 2-25-95,09/23/94, EDG on site fuel supply revised to 4 days e PTSCR 2-25-93,10/08/93, Allow for only one charger out of two in parallel to supply requirements (ref. 3.8.2.3,4.8.2.3.2.c)
e PTSCR 2-15-91,12/18/91, Extended Time Limit on action statement for 120 VAC power from 8 hours to 72 hours e TS Amendm. 177,06/14/94, Paralleling one EDG with 2*' EDG inop e TSCR 2-1-93,08/04/93, Verify EDG auto start on SlAS without loss of offsite power, Amendm.171 e TRMCR 95-2-17,TRM Change Request 3/4.3.2 ESFAS Instrumentation TEMPORARY MODIFICATIONS (Jumpers)
* 2-94-149,12/28/94, LPSI MOV Stem Locking Clamp (Gag Closed) *
e 2-96-084,11/25/96, Temporary Use-As-is Disposition of NCR 296-248, Valve 2-SW-241 Seat Erosion *
* 2-94-022,2/28/92,480V cable penetration in SEP/CRAC room wall e 2-95-016,1/22/95,5 Amp fuses for B EDG potential transformers e 2-96-052, 5/20/96, Provision of a temporary diesel generator e
' 2-93-65,8/11/93, Jumper to bypass Amphenol connector at D RCP Oil Lift Pump *
* 2-92-157,7/15/93, Jumper of alarm contacts to prevent nuisance alarms e 2-94-023, Pwr Hl/Lo Anr'unciator e 2-94-127, ESAS RWST Level * 2-95-146, Pzr Channel Recorder * 2-97-029, RWST'"A" Header instrumentation COMMERCIAL GRADE DEDICATION * CGD 9177700346, CPU Board e CGDF MP2-0148, Woodward Govemor e MP2-0168,10/17/90, Diesel Generator Fuel Injection Nozzle * * 00196260,2" Bronze Valve *
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* MP2-0148,9/26/90, Woodward Governor for Auxiliary Feedwater System Steam Driven Pump Turbine * * MP2-0073,5/30/94, Replacement of non-safety related valve motor operator SMB-1-25 (note 2)*
i MEPL DETERMINATIONS
* MP2-CD-1782,4/11/97, Various Components, Misc, Diesel Generator Components * * MP2-CD-1501,11/21/94, TSP Baskets * * MEPL MP2-CD-1066, CEDS MG Timing Relay TA001 e MEPL MP2-CD-1401, CEDS MG Timing Relay TA-001 e MEPL MP2-CD-1501, RWST Level LS-3002 * MEPL MP2-CD-1680, RWST Level LS-3002 * MEPL MP2-CD-1845, RWST Level LS-3002 e MEPL MP2-CD-683, RWST Level LS-3002 REPLACEMENT ITEM EVALUATIONS (RIES)
e RIE-96-0572,9/26/96, Replacement of non-safety circuit breaker (note 2) *
* Replacement item Evaluation Form RIE-96-0293 5/2/96. *
MISCELLANEOUS LIST
* Parsons Discrepancy Reports DR-0064,0203,0244,308,0341,0342,.0407,0445,0456, and 0523. *
e SQR 2-96-019 " Modification of Control Room Cabinets to Prevent Seismic Interaction and Achieve G.L. 87-02/USI A-46 Outlier Resolution" Rev. * Technical Specification 3.8.2.3, D.C. Distribution - Operatin e Technical Specification 3.8.2.4, D.C. Distribution - Shutdow * NRC Ltr. to NNEC dated 1/23/98, issuance of Amendment Relating to the Emergency Diesel Generators as Described in the Updated Final Safety Analysis Report for the Millstone Nuclear Power Station, Unit 2, and attached SE * EMICE-R10P-01 Rev 1, EMI/RFI Test Report for Johnson Yokogawa Corporation UR1000 Recorders Models 436001,436002,436003, and 436004 e EMF-89-191, Millstone Unit 2 Principal Plant Parameters Notes: 1. (*) Indicates documents reviewed by the NRC and by Parsons 2. Document is for equipment that sentes nonsafety-related functions
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Appendix D List of Acronyms AFW auxiliary feedwater ANSI American National Standards Institute ASME American Society of Mechanical Engineers-CFR Code of FederalRegulations CMP configuration management plan
.CR condition report CST . condensate storage tank ECCS emergency core cooling system EDG emergency diesel generator EMC electromagnetic compatibility EMI electromagnetic interference
- EOP emergency operation procedure ESF engineered safety feature .ESFAS engineered safety feature actuation system l j
EWR engineering work request i FSAR Final Safety Analysis Report FSARCR Final Safety Analysis Report Change Request HELB high-energy line break HPSI high pressure service injection ICAVP Independent Corrective Action Verification Program IFl inspection followup item IST inservice testing requirement LER Licensee Event Report LOCA loss-of-coolant accident MCC motor control center NC nonconformance reports i' NCV non-cited violation NGP Nuclear Group Procedure NNECO Northeast Nuclear Energy Company NRC- U.S. Nuclear Regulatory Commission < PAM post-accident monitoring P&lD piping and instrumentation diagrams - PDCR' Plant Design Change Request PORC Plant Operations Review Committee PTSCR Plant Technical Specification Change Request , D-1
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RWST refueling water storage tank SAR safety analysis report SE safety evaluation SER safety evaluation report SP surveillance procedure SWS service water system TRM Technical Requirements Manual i' TS Technical Specification UIR unresolved item report USQ unresolved safety question VIO - violation
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