IR 05000361/1998014
| ML20198S920 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 01/04/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20198S894 | List: |
| References | |
| 50-361-98-14, 50-362-98-14, NUDOCS 9901120040 | |
| Download: ML20198S920 (52) | |
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ENCLOSURE 2 U.S. NUCLEAR REGULATORY COMMISSibN
REGION IV
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. Docket Nos.: 50-361; 50-362 '
. License Nos.: NPF-10; NPF-15 i
Report No'.: 50-361/98-14;50-362/98-14 !
Liersde: Southern California Edison C Facility: San Onofre Nuclear Generating Station, Units 2 and 3 Location: 5000 S. Pacific Coast Hw l San Clemente, California ,
Dates: October 19 through November 6,1998 .
Team Leader: . Linda Joy Smith, Senior Reactor inspector, Engineering Branch inspectors: David Loveless, Senior Resident inspector, Project Branch C . l
Paula Goldberg, Reactor Inspector, Engineering Dranch Accompanied by: Haywood Anderson, Consuibot Jim Lievo, Consultant Approved By: Thomas F. Stetka, Acting Chief Engineering Branch ]
ATTACHMENT: SupplementalInformation .
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EXECUTIVE SUMMARY I-San Onofre Nuclear Gerierating Station, Units 2 and 3 NRC Inspection Report 50-382/98-14
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During the weeks of October 19 and November 2,1998, the NRC conducted the onsite portion of an engineering team inspection. The team inspection included a review of the design and licensing basis for the component cooling water system and associated support systems and a
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review of the 10 CFR 50.59 safety evaluation program. The team found the overall 4-performance of the engineering organizations to be goo *
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The team found that the design of the component cooling water system was generally c'onsistent with applicable licensing, design, and operations documents. (Section E1).
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Since original construction, appropriate quality standards as described in Updated Final Safety Analysis Report Section 9.2.2.1,"[Cempenent Cooling Water] Design Bases,"
and in Appendix 3.2A, "Q-List," were not specified and included in design documents for j the electrical control of Component Cooling Water Miniflow Control Valves '
2(3) HY-6537, -6538, and -6539. Specifically, the electrical control circuits for these
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valves were not classified as Quality Class ll and as a consequence these control circuits were not provided with Class 1E power and various circuit components, such as
, the controlling flow switches, were not Class 1E or Seismic Category 1. This failure to
!. assure that appropriate quality standards were specified was considered k' be a
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violation of 10 CFR Part 50, AppendL 3, Criterion lil, ' Design Control."(Section E1.2.1). j
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Safety-related setpoints and indications used for Technical Specification compliance ,
adequately accounted for instrument uncertainty. In general, the component cooling i
. water instrument uncertal: ty calculations were retrievable, auditable, thorough,
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consistent with the design and licensing basis, and consistent with the governing instrument calibration procedures (Section E1.2.2).
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The component cooling water analog instrumentation configuration conformed to the
- licensee's Regulatory Guide 1.97 commitments (Section E1.3).
! e' With minor exceptions, the team found the supporting electrical calculations to be retrievable, auditable, thorough, and consistent with the electrical surveillances sampled. In addition, the licensee exhibited a good understanding and sense of
- ownership of the calculations and planned to make needed clarifications (Section E1.4).
, .* Based on the calculations reviewed, the team concluded that the electrical distribution 2 system would adequately support the component cooling water safety functions in
accordance with the design and licensing basis (Section E1.4).
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- The 125 Vdc system direct current distribution Panels D1P1, D2P1, D3P1, and D4P1
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were not conservatively designed to provide optimal protection from a line-to-line bolted
fault near these panels (Section E1.4).
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- One unresolved item was identified concerning control of the thermal performance of the auxiliary salt water / component cooling water heat exchangers. This item is unresolved pending NRC evaluation of the results of planned testing to better estimate the total tube wall fouling of these heat exchangers and NRC review of the licensee's evaluation of past operability (Section E2.2).
- Mechanical, electrical and instrumentation surveillance requirements were found to be appropriately implemented (Sections E2.3, E2.4 and E2.5).
- The licensee's program for managing temporary modifications and partial modifications was acceptable (Sections E3.1 and E3.2).
- Procedural guidance related to the implementation of the requirements of 10 CFR 50.59 was not consistent with current NRC interpretations of the regulation. However, management controls established in the procedure required additional reviews should j the consequences increase above Updated Final Safety Analysis Report limit j Additionally, the safety evaluations reviewed were well documented and properly concluded that no unreviewed safety questions existed ir dicating that the management controls were adequate (Section E3.3).
- The inspectors determined that the licensee had established a reasonable program to identify and address computer components / applications that would not operate properly beyond the millennium. Corrective actions for the specific items reviewed were appropriate and being implemented in a timely manner. The schedule and schedule adherence supported completion of the project prior to December 31,1999 (Section E6.1).
- The licensee recognized an adverse trend with respect to the action request backlog and was developing work management controls to address this issue. All outstanding action requests had recently been reprioritized. The action requests that were reviewed by the team during the course of the inspection were appropriately prioritized (Section E6.2).
- In general, the action request process was acceptably implemented (Section E7.1).
- The failure, during original design and licensing, to assure that applicable design basis parameters for the condensate storage tank were translated into the Technical Specifications and associated surveillance test procedures was in violation of 10 CFR Part 50, Appendix B, Criterion lil, " Design Control." This nonrepetitive, licensee identified and corrected violation is being treated as a noncited violation consistent with Section Vll.B.1 of the NRC Enforcement Policy (Section E8.1).
- In 1996, due to a word processing error, the licensee failed to assure that applicable design basis parameters for the condensate storage tank were translated into operating procedures. This nonrepetitive, licensee-identified and corrected violation of 10 CFR Part 50, Appendix B, Criterion 111," Design Control,"is being treated as a noncited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy (Section E8.2).
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An incorrect spring installed in the Unit 3, Train A, component cooling water surge tank pressure relief valve 3PSV-6356, resulted in Train A of component cooling water being inoperable from March 2,1987 to February 21,1992, in violation of Technical Specification 3/4.7.3. In 1983, the valve vendor had provided the licensee with an incorrect spring part number for this valve, which was maintained in inventory until its use in 1987. This nonrepetitive, licensee-identified and corrected violation is being treated as a noncited violation consistent with Section Vll.B.1 of the NRC Enforcement Policy (Section E8.3).
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. From December 1,1993, to April 27,1997, a lift stop of the wrong length was installed in the Unit 3, Train A, component cooling water surge tank pressure relief Valve 3PSV-6356. As a result, Valve 3PSV-6356 did not provide the overpressure protection specified in ND-7000 of1974 Edition of Section lli of the ASME Boiler and Pressure Vessel Code, in violation of 10 CFR 50.55a(a)(2). The licensee determined that at the time of procurement, the vendor had failed to provide instructions for modifying the length of the lift stop, and it had been installed as received. This nonrepetitive, licensee-identified and corrected violation is being treated as a noncited violation consistent with Section Vll.B.1 of the NRC Enforcement Policy (Section E8.3).
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Table of Contents
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li t . Enginee ring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 El Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 E1.1 Component Cooling Water Mechanical System Design . . . . . . . . . . . . 1 E1.2 Component Cooling Water instrumentation and Control System Design ;
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E1.2.1 Implementation of Component Cooling Water Control System Functional Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 ;
E1.2.2 Component Cooling Water Setpoints and instrument Uncertainty 6
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E1.3 Regulatory Guide Compliance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 E1.4 ' . Component Cooling Water Electrical System Design . . . . . . . . . . . . . . 8 j E2 Engineering Support of Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . 10
' E2.1 Component Cooling Water System Walkdown . . . . . . . . . . . . . . . . . . .10 ;
E2.2 Component Cooling Water Heat Exchanger Performance Testing . . . .- 10 E2.3 Component Cooling Water (CCW) Mechanical System Surveillance
- Testing Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 E2.4 Component Cooling Water Instrument Calibration . . . . . . . . . . . . . . . 14 E2.5 Component Cooling Water Electrical System Surveillance Testing Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 E3 Engineering Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . . 14 E3.1 Tampa : / Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 E3.2 Partial Modifice.tions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 E3.3 Review of the 10 CFR 50.59 Safety Evaluation Program . . . . . . . . . . 16 :
E6 Engineering Organization and Administration . . . . . . . . . . . . . . . . . . . . . . . . 19 E6.1. Computer Compliance for the Year 2000...................... 19 E6.2. Engineering Backlog . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 - !
E7 Quality Assurance in Engineering Activities . . . . . . . . . . . . . . . . . . . . . . . . . . 21 E7.1 Action Req uests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 -!
E (Closed) Licensee Event Report 50361/96-012: Condensate Storage Tank Outside its Design Basis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 E8.2 (Closed) Licensee Event Report 50-361/98-009-01: Condensate Storage Tank Outside Design Basis Due to Procedural Error . . . . . . . . . . . . . 26 E8.3 (Closed) Licensee Event Report 50-362/98-012-00 and -01: Componen ,
Cooling Water Relief Valve Inoperable - Wrong Set Spring Installed Due to Vendor Data Error . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 E8.4 (Closed) Licensee Event Report 50-361/98-014: Emergency Feedwater Actuation Signal Outside Design Basis . . . . . . . . . . . . . . . . . . . . . . . . 29 V. Maaagement Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 X1 Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . '29-v-a
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., 1 4-Reoort Details jnspection Objectives This inspection was performed in accordance with two core inspection procedures: " Safety
' System Engineering Inspection, " (93809) and "10 CFR 50.59 Evaluations," (37001). The team reviewed the design basis documentation for the component cooling water (CCW) syste This system was selected because of its relatively high risk significance. The team also evaluated implementation of the safety evaluation progra Ill. Enaineerina E1 Conduct of Engineering (93809)
E Component Coolina Water Mechanical System Desian Insoection Scooe The team reviewed various CCW calculations and compared them to the available licensing, design and operations documents related to the capability of the CCW system, including the CCW safety-related makeup syste Observations and Findinas Component Cooling Water Pump / Component Cooling Water Makeup Pump Net Positive Suction Head The team found that the available net positive suction head for the component cooling water pump and the associated component cooling water make up pump exceeded their respective required net positive suction hea Available Primary Plant Makeup Storage Tank Capacity The team determined that the licensee had appropriately determined required primary plant makeup storage tank capacity versus CCW system leakage and had correctly specified the total allowable CCW leakage versus primary plant makeup tank level in Technical Specification Table 3.7.7.1- Conclusions The team concluded that the CCW pump and the component cooling water make up pump were both provided adequate net positive suction head. The licensee had also
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correctly specified the total allowable CCW leakage versus primary plant makeup tank level in Technical Specification Table 3.7.7.1- ,
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b E1.2 Component Coolino Water Instrumentation and Control System Desian E1.2.1 Implementation of Component Cooling Water Control System Functional Requirements Inspection Scoce The purpose of this review was to confirm that functional requirements of the Updated Final Safety Analysis Report and CCW Design Basis Document had been adequately implemented in the design, including safety classification, single failure, and independence / separation requirements. The team reviewed Updated Final Safety Analysis Report Section 9.2.2, " Component Cooling Water System," Revbion 13; Design Basis Document DBD-S023-400, " Component Cooling Water," Revision 5; loep diagrams for CCW surge tank level, CCW flow, CCW discharge pressure; and elementary diagrams for the CCW pumps, CCW miniflow control valves, non-critical loop isolation valves, and CCW containment isolation valve Observations and Findinas For the control circuits reviewed, with the exception of the miniflow control valves, the team found that the circuits would perform their safety functions in accordance with the licensing and design basi Component Cooling Water Miniflow Control Valves Miniflow Control Valves 2(3) HY-6537, -6538, and -6539 were designed to open below 3000 gpm CCW flow and close above 3000 gpm. The licensee stated that they did not take credit for the pump miniflow protection function because the expected flow paths during an engineered safety features actuation were sufficient for pump protectio These control valves were configured to close on loss of air or control power. In each unit, the three miniflow control valves were powered from a single nonsafety-related battery. As such, they could be expected to operate as designed following a i loss-of-offsite powe The licensee stated that, in general, their original license basis was that non-Class 1E 1 devices that are not required to change position or operate during or after a design j basis event were considered to lose power at the initiation of the event and fail in their I safe position. As a result, the failure modes and effects analysis developed in the l original Final Safety Analysis Report were based on the assumption that the miniflow !
control valves would lose their non-Class 1E power and fail in the safe position, close This assumption was important because these valves needed to be closed to prevent diversion of CCW flow, which would prevent CCW from providing required cooling for various safety-related thermalloads needed for the mitigation of design basis accidents or for safe shutdow In 1988, as a result of an NRC safety system functionalinspection, the licensee recognized that their original license basis was flawed, in that non-Class 1E devices I could spuriously actuate, especially during an earthquake. They recognized the need to l reevaluate non-Class 1E devices to ensure that for all credible events, they met 10 CFR 50, Appendix A, General Design Criterion 2 (GDC 2)," Design Bases for-2-
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Protection Against Natural Phenomena," and 10 CFR Part 50, Appendix A, General Design Criterion 44 (GDC 44)," Cooling Water." This issue was reported to the NRC in Licensee Event Report 50 361/88-034, The results of the licensee's full evaluation of the issue, which was docu mented in Report No. M86420, " Spurious Actuation Evaluation, Component Coc5ng Water System Operability Assessment," dated February l
1990, were forwarded to the NRC on March 7,1990. The NRC did not prepare a safety '
evaluation for this submittal or document review of this repor Af ter making some initial modifications, performing seismic inspections and doing comparisons with similar seismically qualified devices, the licensee concluded that limited use of non-Class 1E devices had only a minor effect on CCW system operation and did not adversely affect the CCW safety function. They further stated that a thorough review of the non-Class 1E devices using the methodology established in Generic Letter 87-02," Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors (USl A-46)," demonstrated that the CCW system at SONGS 2 & 3 complies with GDC 2 and GDC 4 During a drawing consolidation effort in 1993, the licensee again questioned the quality i classification of the component cooling water miniflow control valves and identified a l potential common mode failure of all three valves to their op n position. The licensee initiated Nonconformance Reports 930600044 through 9306v 149 dated July 1,1994 to i evaluate the effects of postulated high impedance faults in circ .s common to raceways shared by the miniflow control circuits that would not have been cleared by existing protective devices. The licensee performed a circuit failure analysis (A-94-E-001, CCW Miniflow Valves Circuit Analysis, Revision 0, September 27,1994) evaluating these faults (called " toaster wire" failures by the licensee). Backup fuses were installed to provide redundant electrical protection, which would preclude spurious opening of the valves by this mechanis The team noted that in both of these previous reviews, the licensee had not correctly analyzed the safety function of the miniflow control valves to close. Licensee Event Report 88-034, the failure modes and effects analysis in Report M86420, Calculation A-94-E-001, and the 10 CFR 50.59 safety evaluation for Field Change Notice F11676M (which restored power to the miniflow control valve circuits) all incorrectly assumed that the valves would not change position during accident or safe shutdown operation, and that the valves themselves were safety-related for a pressure boundary function onl The team found that the CCW system was designated in Updated Final Safety Analysis Report Section 7.1 as an auxiliary support system for the engineered safety features that are provided to mitigate the consequences of postulated accidents. During a design basis accident with a loss-of-offsite power, these valves are designed to open and must close af ter the CCW pumps are restarted to enable the CCW system to deliver required design basis flows. The team determined that CCW Miniflow Control Valves 2(3) HY-6537, -6538, and -6539 perform a safety action to close, which is necessary to mitigate the consequences of postulated accident .
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d Updated Final Safety Analysis Report Section 9.2.2.1,"[ Component Cooling Water)
Design Bases," stated that the component cooling water system and components are designed to the equipment classification requirements indicated in Append 5 3.2A, "O-List." Appendix 3.2A defined Quality Class 11 as those strucures, components, and systems not in Quality Class I that are provided to mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the publi Appendix 3.2A stated that Quality Class 111 was for those components not in Quality Class I or ll, whose limited damage could interrupt power generation or release radioactive materials to the environment in excess of average release limits, but not in excess of licensed release limit The miniflow control valves and the associated air-operator, solenoid valves, backup air accumulators, flow transmitters and low flow annunciators were all designated as Quality Class 11, Seismic Category I. However, the signal conditioners and flow switches used for opening and closing the valves, and the electric power supply were designated as non-Class 1E (Quality Class ill, Seismic Category ll). The three miniflow valves (per unit) shared a common service breaker from the non-Class 1E direct current system with additional Quality Class 111 radwaste circuits. The miniflow valve electrical control circuits were not independent or separate In reviewing the Updated Final Safety Analysis Report, the licensee's past evaluations, and the present design configuration, the team also noted that the Updated Final Safety Analysis Report did not include the miniflow control valves in the failure mode and effects analysis provided in Updated Final Safety Analysis Report Table 9.2-3. The team concluded that since these valves received battery power and since these valves were designed to open following a loss-of-offsite power, the licenseo should have evaluated the failure of these valves to close on demand, and consequent?y should have classified these circuits as Qualitf Class ll, Seismic Category 1, rather than Quality Class 111, Seismic Category 1 Classification as Quality Class ll would require the licensee to design the circuits in accordance with IEEE Std 279-1971 criteria for single failure, quality, separation, independence, indbation of bypass, and information read-out. Assuming credit for the licensee's seismic interaction evaluations submitted to NRC and the corrective actions taken for precluding " toaster wire" failures within raceways, the current configuration still did not satisfy these criteria for the following reason * The circuits were not separated, and shared the same terminal blocks and enclosures; Updated Final Safety Analysis Report Section 8.3.3.3.3 stipulates that different separation groups must be separated by six inches of air space or else barriers or enclosed raceways shall be used. There were several opportunities for short circuiting adjacent terminais and spuriously opening the miniflow valve (s), without creating a f aulted condition that would be mitigated by the circuit protective devices; there was also a potential for spurious operation resulting from postulated multiple direct current grounds. No analysis of these types of f ailures resulting from lack of separation at cable terminations or within cabinets was evident to the team.
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- The common, nonsafety-related, ungrounded direct current system would not be expected to have the degree of maintenance scrutiny and urgency for identification, isolation, and repair of grounds on the system as for a safety-related system. The existence of a ground on the system can " arm" the circuits such that a subsequent ground could spuriously open a miniflow valv Subsequent multiple grounds could be masked if the first ground was not isolated in a timely manner; this could result in spurious opening of more than one m miilow valve, because they share the same batter * Paragraph 4.13 of IEEE Std 279-1971 required in part that,"If the protective action of some part of the system has been L fpassed or deliberately rendered inoperative for any purpose, this fact shall be continuously indicated in the control room." There was no remote position indication or remote control provided for the miniflow control valves; therefore, if the valve were fully or partially open, there would be no indication of this bypassed condition in the control room, and no means to correct the condition in the control roo In response to the team's concern, the licensee issued Action Request 981100386 dated November 5,1998 which committed to interim removal of control power from the miniflow control valve solenoids, thereby assuring that CCW flow would not be unintentionally bypassed from its design basis heat loads. The licensee concluded in this action request that, based on Repod M86420 and Calculation A-94-E-001,"the valves and their control system components are satisfactory to perform their safety function," and implied that the primary concern was that the licensee's ". . level of documentation of the valves' active safety function was lacking."
Notwithstanding the licensee's position, the team concluded that with power connected, the valve comrol circuits were not adequate for performing their safety function, because the circuits did not conform to IEEE Std 279 criteria and Updated Final Safety Analysis Report commitments for single failure, separation, independence, indication of bypass and information read-out. This nonconformance to design and licensing basis criteria for safety-related circuits was a consequence of the licensee's failure to recognize the active safety function of the valve circuits and failure to classify them as Quality Class 1 CFR Part 50, Appendix B, Criterion 111. " Design Control," states, in part, " Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 6 50.2 and as specified in the license application . . . are correctly translated into specifications, drawings, procedures, and instructions. These measures shall include provisions to assure that npropriate quality standards are specified and included in design documents. " Tho .ailure to assure that appropriate quality standards were specified and included in design documents for the electrical control of the component cooling water miniflow control valves is a violation of NRC requirements (50-361/9814-01;50-362/9814-01).
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Configuration Control of Evaluated Non-Class 1E Components Subsequent to the completion of the inspection, the team asked the licensee how they
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assured appropriate quality standards for the non-Class 1E components that had been evaluated as capable of withstanding a seismic event in Report M86420, which was discussed in the previous section. The licensee was not initially able to identify any
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method for assuring that the current configuration matched the configuration that was evaluated in Report M86420. They initiated an action request to perform a formal investigation of this issue. Inspection follow up is planned to evaluate the licensee's review (50-361/9814-02; 50-362/9814-02).
Minor Discrepancy in Component Cooling Water Design Basis Document Regarding Miniflow Line The licensee stated that they did not take credit for the miniflow protection function because of the other flow paths available in the system. However, Design Basis Document DBD-SO20-400, Section 4.1, stipulated in part that:
l "To achieve its safety function, a [CCW] pump must meet the following requirements:
" . . minimum flow lines for flow below 3000 gpm . . . ."
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The team concluded the foregoing description was not clear, because it suggested that credit was required for the " minimum flow lines." The licensee explained that this section of the design basis document was intended to specify the component requirements for the pump, and not the system configuration required to protect the
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. Conclusions I For the control circuits reviewed, with the exception of the miniflow control valves, the team concluded that the circuits would perform their safety functions in accordance with the licensing and design basis. Regarding the exception, the electric control circuits for
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the component cooling water miniflow control valves were not appropriately classified as
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Quality Class ll and, as a consequence, the circuits were not separated, seismically ;
qualified and were not provided with either control or bypass indication in the control room. This failure to assure that appropriate quality standards were specified and included in design documents for the electrical control of the component cooling water 4 miniflow control valves is a violation of 10 CFR Part 50, Appendix B, Criterion lit,
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" Design Control."
E1.2.2 Component Cooling Water Setpoints and Instrument Uncertainty
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The team reviewed eleven instrument uncertainty calculations for setpoints or indication
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channels involving the CCW system. Channels that were safety-related or important to safety were sampled. The attributes reviewed for the calculations included scope,
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assumptions, design inputs, validity of the design / licensing basis and methodology; consistency of selected interface requirements (for example, with other calculations or drawings and modifications); and identification of design margin Observations and Findinos With the exception of a document discrepancy identified in calculation J HEA-010, the l team found the CCW instrument uncertainty calculations to be retrievable, auditable, thorough, and consistent with the design and licensing basis, and with the instrument calibration procedures sample Scaling Calculation J-HEA-010 Tables 1 and 2 contained a footnote stating, in part, that l
" Unrecoverable [PPMST) volume for CCW makeup use is 10575 gallons." However, the i setpoint calculation J-HEA-011 and Calculation M-0027-020, Rev. O,"CCW Safety Related Makeup System - Hydraulic Calculations," identified an unrecoverable volume of 11,011 gallons. The licensee determined that the value shown in the scaling calculation was incorrect, and stated an action request would be prepared to correct Calculation J HEA-010. The team concluded that the error in the scaling calculation did not affect the result of the scaling calculatio Conclusions Based on this review, the team concluded that the licensee had adequately accounted for instrument uncertainty for safety-related setpoints and indications used for Technical Specification compliance. In general, the CCW instrument uncertainty calculations were retrievable, auditable, thorough, consistent with the design and licensing basis, and consistent with the governing instrument calibration procedure E1.3 Reaulatory Guide Comoliance Insoection Scope The team reviewed the licensee's commitment to Regulatory Guide 1.97,
" Instrumentation for Light Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Revision 2, for CCW flow (2FT-6243/6248) and CCW temperature (T6469/6470), and reviewed the loop diagrams and equipment data to evaluate compliance with the licensee's commitmen Observations and Findinas The team found that the licensee had implemented these loops in accordance with their commitment to Regulatory Guide 1.97 Category 2 Type D variables for the CCW system with respect to quality classification, range, equipment qualification, and power source Conclusions The team concluded that the CCW analog instrumentation configuration conformed to the licensee's Regulatory Guide 1.97 commitmen I .
E1.4 - Comoonent Coolina Water Electrical System Desian Insoection Scope To confirm the adequacy of alternating current electrical support for CCW equipment, including consideration of degraded voltage and protective relaying, the team reviewed a sample of electrical calculations for the safety-related 4160 Volt and 480 Volt alternating current systems serving CCW loads. The team reviewed the governing calculations for the diesel generator t,teady-state and transient loading, alternating current voltage regulation, and selected electrical protection of the alternating current load path serving CCW Pump 2P-024 and CCW motor-operated Valve 2HV-6236 (CCW non-cntical containment outlet isolation valve).
To confirm the adequacy of direct current electrical support for CCW equipment, the i
team reviewed the direct current control circuita for 4160 Vac Bus 2A04. Two calculations were reviewed to evaluate the adequacy of direct current voltage and protection for this load pat The attributes reviewed for the calculations included scope, assumptions, design inputs, validity of the design / licensing basis and methodology, consistency of selected interface i requirements (for example, with other calculations or drawings and modifications), and ,
identification of design margin ' Observations and Findinas With minor exceptions, the team found the supporting electrical calculations to be retrievable, auditable, thorough, and consistent with the electrical surveillances sampled i (See Section E2.5) In addition, the licensee exhibited a good understanding and sense of ownership of the calculations and planned to make needed clarification Alternating Current Circuits For the CCW alternating current load paths sampled, the calculations demonstrated adequate diesel generator capacity and transient loading capability; adequate voltage regulation under design basis conditions, including degraded voltage; adequate electrical protection of equipment and conductors; and adequate equipment rating Direct Current Circuits For the CCW direct current load paths sampled, the calculations demonstrated adequate voltage regulation under design basis conditions, including degraded voltag The team noted that IEEE 308-1974, "lEEE Standard Criteria for Class 1E Electric Systems for Nuclear Power Generating Stations," specified that protective devices shall be provided to limit the degradation of the Class 1E power systems. The licensee's commitment to IEEE 308-1974, which was described in Updated Final Safety Analysis Report Section 8.3.2.2.1.7, stated, " Equipment of the Class 1E de system is protected and isolated by fuses or circuit breakers in case of short circuit or overload conditions."
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In Calculation E4C-109, " Class 1E 125 Vdc System Protection Calculation," dated June 17,1993, and Calculation Change Notice CCN-5, dated June 14,1996, the licensee identified a potential deviation from the licensee's commitment to IEEE-30 The licensee identified that the available short circuit current at direct current distribution ,
Panels D1P1, D2P1, D3P1, and D4P1 for a kne-to-line bolted fault would exceed the l
10,000A withstand rating of the panels and breakers, and exceed the breakers' 5000A l interrupting rating. The licensee accepted this condition of the distribution panel I breakers as-is, because the calculation concluded that the vulnerability to bolted faults I that would theoretically excoed the ratings would be limited to a maximum length of about 34 feet of cable. That these cable segments were located in a mild environment and were not exposed to missiles or high energy line break hazards that could cause such a fault. The licensee determined that the only credible cause of line-to-line bolted j fault within 34 feet from the panel was a fire. They noted that this condition was addressed in the Appendix R analysis as acceptable in that fires are not postulated in ,
both Class 1E 125 Vdc distribution rooms simultaneously. If a fault occurred,it would I be isolated by the upstream feeder breaker from the respective battery bus; and that even if the fault was not isolated, loss of the entire 125 Vdc system division would be an event and consequence bounded by the plant safety analysis. As an additional mitigating factor, the licensee noted that simultaneous maintenance in more than one direct current distribution pa.1e! room was prohibited by procedur l
!! was the licensee's position that the current configuration complied with IEEE 308-1974 in part, because IEEE 308-1974 does not contain explicit requirements as to how to apply short circuit analysis for the protective devices. They concluded that the Class 1E l 125 Volt DC panel breakers were acceptable under all credible short circuit condition They stated that at the time San Onofre was designed (1973 to 1978), the electrical design followed standard industry practices for breaker and panel sizin The team accepted the licensee's position in part. The team agreed with the licensee's engineering evaluation of this condition and with the conclusion that there would be a low probability of such a fault. However, the team noted that isolation of a design basis fault (line-to line bolted fault) in the potentially vulnerable part of the circuit would rely on upstream protection. As a result, the entire 125 Vdc distribution panel would be isolated rather than limiting the degradation to the faulted branch circuit. The team noted
, that in the early 1970s, manufacturers provided equipment with withstand ratings and interrupting ratings so that the designer could select breakers and panels that were sized for the limiting short circuit current. The team determined that the licensee had not conservatively implemented IEEE-308-1974 and Updated Final Safety Analysis Report Section 8.3.2.2.1.7, in that the breakers in direct current distribution Panels t D1P1, D2P1, D3P1, and D4P1 did not provide adequate protection for the theoretically limiting short circuit. However, the team acknowledged that the current design would adequately protect against the short circuit currents expected for credible faults and concluded that this design weakness did not affect equipment operabilit .. .. _- - . - - . - _ _ .
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. Conclusions
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With minor exceptions, the team found the supporting electrical calculations to be retrievable, auditable, thorough, and consistent with the electrical surveillances sampled. in addition, the licensee exhibited a good understanding and sense of ownership of the calculations and planned to make needed clarifications.
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Based on the calculations reviewed, the team concluded that the alternating current electrical distribution system would adequately support the CCW safety functions in accordance with the design and licensing basis.
Based on the calculations reviewed, the team concluded that the 125 Vdc system would adequately support the CCW safety functions in accordance with the design and licensing basis. However, one weakness was identified in that direct current distribution Panels D1P1, D2P1, D3P1, and D4P1 were not conservatively designed to provide optimal protection from a line-to-line bolted fault near these distribution panels.
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E2 Engineering Support of Facilities and Equipment (93809)
- E Component Coolina Water System Walkdown
4 The team performed a walkdown of the CCW instrumentation and electrical areas. This
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included the surge tank instrumentation, CCW flow transmitters 2FT-6243/6248, the
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CCW process radiation monitor; miniflow control valves, analog cabinet 2L158FR,4160 Vac switchgear rooms, the transfer switch serving CCW 2P-025, direct current distribution panel rooms, and the front and interior of Main Control Board Panel 2CR064. The team did not identify any problems in these area E2.2 Component Coolina Water Heat Exchanaer Performance Testina
' Inspection Scope The team reviewed the licensee's commitments to Generic Letter 89-13," Service Water System Problems Affecting Safety-Related Equipment," and the implementing procedures and calculations. The team interviewed licensee personnel responsible for heat exchanger testing and analysis. The team also reviewed related analysis, test reports, and procedure Observations and Findinas The team found that on March 29,1991, the licensee committed to performing
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functional testing of the CCW heat exchangers to verify the heat removal rate of the heat exchangers and to trend any degradation in performance.
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! 1 The licensee stated that thermal performance of the component cooling water / auxiliary i j salt water heat exchanger was affected by the total tube wall fouling factor (or R, the i combined inside and outside tube wall resistance) and macro-fouling such as seaweed l and seashell accumulation on the auxiliary saltwater side of the heat exchanger. Both of l these factors were considered in the licensee's program to assure operability of the component cooling water / auxiliary salt water heat exchanger l
!
The licensee's process for assuring heat exchanger operability was based in part on heat exchanger performance testing to determine the fouling factor. Prior to this testing,
- the licensee backflushed the heat exchangers to minimize seaweed and debris l l accumulation. The team agreed that this was acceptable, because the licensee I
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accounted for macro-fouling analytically in Calculation M-0027-023,"CCW Heat Exchanger Operability, Revision 0 through CCN- In Calculation M-0027-023, the licensee calculated predicted heat transfer as a function of heat exchanger delta pressure (a measure of seaweed and debris accumulation),
ocean temperature and auxiliary saltwater flow and developed graphs to define j acceptable regions of operation. These graphs were included in Operating instruction SO23-2-8, " Saltwater Cooling System Operation," Revision 18, and the i operators used them to determine when it was necessary to backflush the heat '
exchangers to assure operabilit During the inspection, the licensee identified that Calculation M-0027-023 was flawed, because it was based on an assumed total tube wall fouling factor, which was less conservative than the most recently measured total tube wall fouling factor for Heat l Exchangers 2ME002,3ME001 and 3ME002. The licensee determined that Calculation M-0027-023 had not been updated to consider the results of testing performed prior to Refueling Outage 9, and on October 28,1998, initiated Action Request 981001913 to address this issue. The Action Request Committee's initial operability review concluded that the heat exchangers were operable based on current l low ocean temperatures and conservatisms built into the calculation. On November 5, 1998, the licensee documented a more complete operability determination, and determined that the most recently measured total tube wall fouling factors were iikely not accurate, because the thermal performance test data was collected at low heat loa They also determined that the higher total tube wall fouling factors measured prior to Refueling Outage 9 did not reflect a change in heat exchanger performance. The licensee noted that test data for Heat Exchanger 2ME001, taken prior to Refueling Outage 8, showed that the total tube wall fouling factor increased as load decrease Total Tube Wall Fouling Factor R, was 6.0 at low load (25 million Stu's) and 5.6 at higher loads (146 million Btu's)." On this basis, they neglected the data taken prior to Refueling Outage 9, and concluded that the existing program for controlling the accumulation of seaweed and debris in the heat exchangers was acceptable; therefore, the heat exchangers were operabl The team agreed that thermal performance data collected at low load was generally less accurate than data collected at high heat load due to the larger absolute impact of instrument error. However, to fully evaluate the acceptability of the licensee's conclusion, the team requested a history of test results. The licensee provided the following information:
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Total Tube Wall Fouling Factor (Hr-sqft-F/ Btu)
- Source of Data ' Heat Exch.an_gg Unit 2 Unit 3 ~ j Calculation ' N/A~ 7E-4 1.1 E-3 -
- M-0027-023 j
"
. Assumption
' Prior to Refueling ME001 ' - 5.6E-4 ' 1.01 E-3 -
Outage 8' High j
~
y
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Heat Load 1 J
- Prior to Refueling ME002; 6.5E-4 8.5E-4 E ' Outage 8 - High
_
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Heat Load Prior to Refueling ME001 ~ 6.0E-4 Not Performed ;
Outage 8 - Low ,
, Heat Load - l I' Prior to Refueling - ME002 Not Performed Not Performed
' Outage 8 - Low Heat Load Prior _to Refueling . ME001 6.7E-4 1.46E-3
.@ 'c
' Outage 9 - Low .
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P' ; Heat Load l
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. Prior to Refueling ME002 8.6E-4 1.47E-3
' Outage 9 ; Low Heat Load
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The team noted that data for Unit 2 Heat Exchanger ME001 collected prior to Refueling Outage 8, during high and low heat load testing, only differed by 0.4E-4; whereas, the .
worst case (Unit 3 Heat Exchanger ME002) difference between prior to Refueling -
Outage 9 low load fouling factor and the prior to Refueling Outage 8 high load fouling ,
'
factor was 6.2E-4. The team was concerned that the prior to Refueling Outage 9 test
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. data could represent a real decline in heat exchanger performance, particularly for the cUnit 3 heat exchangers. To address this concern, the licensee reevaluated their basis q
+ ' for operability. . After considering other margins available in the calculation, the licensee .j 1 developed a new interim set of operability curves for the Unit 3 heat exchangers, based !
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7 on the worst case total tube wall fouling factor measured prior to Refueling Outage !
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-The team concurred with this more conservative operability determinatio ' The . team noted that Engineering Procedure S023-V-3.25," Component Cooling Water-Heat Exchanger Testing," Revision 3, did not include guidance concerning trending .
degradation in heat exchanger performance as indicated in the Program Response for Generic letter 89-13, and as specified in Calculation M-0027-023, Revision 0 through CCN-1. The licensee agreed that the on-line test procedure and test failed to include-12-s
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. l the trending required in the original design calculation. They planned to resolve this issue by performing a retest on two of the four CCW heat exchangers during the upcoming Unit 2 and Unit 3 outages. This retest will include both low and high heat load tests and an evaluation of the total tube wall fouling factor, R., They planned to use this data to evaluate the past operability of the heat exchangers. The failure to update Calculation M-0027-023 to address thermal performance data and the impact on past operability is unresolved. This item is unresolved pending NRC evaluation of the results of the planned heat exchanger performance testing and the licensee's past operability determination (50-361/9814-03; 50-362/9814-03). Conclusions Component cooling water heat exchanger thermal performance testing performed prior to Refueling Outage 9 resulted in total tube wall fouling factors for three heat exchangers that were non-conservative to those assumed in the calculation used by the licensee to determine when backflushing was necessary to control macro-fouling, such ,
,
as seaweed or debris. The licensee planned additional heat exchanger thermal performance testing to accurately measure the total tube wall fouling factor. This item is unresolved pending NRC evaluation of the results of the planned testing and NRC review of the licensee's past operability determinatio E2.3 Component Coolina Water (CCW) Mechanical System Surveillance Testino Acceptance Criteria Inspection Scope The team reviewed supporting calculations and implementing procedures for the following CCW Technical Specification Surveillance Requitements: 3.7.7.1, backup nitrogen supply verification; 3.7.7.4, pump inservice testing; 3.7.7.6, setpoint for the third stage pressure regulator of the backup nitrogen system; and 3.7.7.13, CCW leakage measurement. In addition, the team reviewed the implementing procedure for Technical Specification Surveillance Requirement 3.7.7.2, flow path verification, and recent pump test records related to the inservice test progra Observations and Findinas The team found that the acceptance criteria in the surveillance requirements and the implementing procedures were consistent with the design basis calculation The team confirmed that appropriate flow path valves were included in the surveillance procedure for flow path verification. The team also confirmed that the most recent pump inservice test results were consistent with the pump test acceptance criteria, Conclusions Five mechanical CCW Technical Specification surveillance requirements were found to be supported by appropriate calculations and implementing procedure o.
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< E2.4 Component Coolina Water Instrument Calibration
- Insoection Scoce The team reviewed a sample of calibration procedures and results for the last two l
- calibrations. The review included an evaluation of consistency with respect to corresponding instrument uncertainty calculations and adequacy of the recent
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performance history for the instrument. Instruments sampled were CCW Flow Transmitters 2FT 6243/6248 and surge tank level Transmitter 2LT-6498. Calibration of differential pressure transmitters was governed by Procedure SO23-II-9.14,
, " Electronic Differential Pressure and Pressure Transmitter Calibration," Revision 1. The team retrieved calibration history from the licensee's MOSAIC data bas l
, Observations and Findinas The calibration tolerances, measuring and test equipment, and procedures were consistent with the corresponding calculations. The calibration histories were
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acceptable, and when out-of-tolerance conditions were found, they were adequately
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dispositioned. The six instrument calibrations that were sampled were acceptabl E2.5 Component Coolina Water Electrical System Surveillance Testina Acceptance Criteria Inspection Scope
The team reviewed a sample of three electrical surveillances and their results to evaluate consistency with the Technical Specifications and with the supporting electrical
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calculations.
- Observations and Findinas
Based on a programmatic review of the surveillances, the team found them consistent i with Technical Specification requirements and corresponding electrical calculation Disposition of data from the last surveillances performed to the procedure appeared
, acceptable. The three electrical surveillances that were sampled were acceptabl E3 Engineering Procedures and Documentation E Temporary Modifications (37550) inspection Scope The team reviewed the licensee's temporary modification program and discussed selected safety related temporary modifications with appropriate licensee personnel. In j addition, the team reviewed Procedure SO123-XV-5.1," Temporary Modification I Control," Revision 2. The team discussed the temporary modifications and the procedure requirements with licensee personne l
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! Observations and Findinas l l
The team noted that the procedure provided three methods for performing a temporary modification: facility modification, nonconformance report, or a maintenance orde Facility modification packages were prepared, if the modification impacted operating procedures or drawings in the centrol room drawing file. A nonconformance report was I used for modifica; ions with only minor impact on operating procedures or control room file drawings. Maintenance orders were used for short term nonsafety-related system / equipment repairs that did not affect control room file drawings or operating procedures. At the time of the inspection, there were only two temporary facilities modifications and only one of them was safety-related. The team reviewed the two temporary facilities modifications and found them acceptable. The team also reviewed 17 temporary modifications and found them to be acceptabl l l Conclusions Based on these reviews, the team concluded that the licensee was effectively controlling temporary modifications to the facilit ,
E3.2 Partial Modifications (37550) l inspection Scope j The team reviewed selected modifications which had been partially implemented and l discussed their status with licensee personnel. The team also reviewed the results o' l the licensee's 1996 self-assessment, which addressed partial implementation of modification Insoection Foitow uo in the 1996 self-assessment, the licensee identified that there were approximately 150 open design change packages. Some of these packages had been open since 198 During this inspection, the team reviewed the licensee's design change package closure matrix, dated October 20,1998, which contained a list of 121 design change packages and their status. The team found that the licensee had established four different categories for tracking partialimplementation of modifications: Design change packages that were engineered but not implemented in the field (this category had a total number of 27 design change packages with 10 of them remaining open). Design change packages that were partially implemented in the field and the equipment had been turned over to the operations staff, but the modification had not been formally closed (there were 20 in this category with six remaining open).
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- . Design change packages that were completely implemented in the field and the equipment had been turned over to the operations staff, but the modification had not been formally closed (there were 70 in this category with 16 rernaining open).
1 Design change packages that had a work-in-progress status (there were four in 1- this category).
The team reviewed three modification packages from the first category and found that
, the modifications were completed on Unit 2 and canceled on Unit 3. The team discussed the modifications with the licensee and determined that the modifications i
) canceled on Unit 3 were not necessary and had no safety significanc I
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I l The team reviewed three modification packages for the second category and found that j the modifications were fully implemented on Unit 2 and canceled on Unit 3 or were fully !
implemented on both Units with some of the final paperwork not complete. Again, the team determined that, for the modifications that were canceled on Unit 3, the modifications were not necessary and had no safety significance. The team determined l that, for the modifications that were implemented in both Units, the incomplete paperwork was not safety significan The team reviewed four packagts from the third category. The team determined that missing documentation consistec of the final sign off on a sheet verifying that all required documents were completed. The team noted that the document changes were made years before the final verification was completed. Again, the team did not find any safety significant item The items in the fourth category involved preplanning for forced outage work, such as performing the engineering necessary to replace a reactor coolant pump, if neede Conclusions The team concluded that, while the licensee's process for reviewing and closing modifications was slow, the licensee was adequately controlling partial implementation of modification E3.3 Review of the 10 CFR 50.59 Safety Evaluation Prooram (37001) Inspection Scope The team reviewed components of the licensee's unreviewed safety question determination program in accordance with Inspection Module 37001. Several formal safety evaluations and screening evaluations were reviewed. Procedural guidance for implementing safety evaluations, updating the Final Safety Analysis Report, and maintaining related records was reviewed. General Engineering Design Procedure SO123-XV 44," Guidelines for Determining When a 10 CFR 50.59 Safety Evaluation is Required," Revision 2, and the associated Form 26-548, " Engineering Design Program 10 CFR 50.59 Safety Evaluation," Revision 3, were reviewe i l
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O b. Observations and Findinas The team reviewed the programmatic procedures to ensure that the formal guidance provided in Procedure SO123-XV-44 properly implemented the requirements of ,
Section 50.59. The guidance included assessing and documenting whether l Section 50.59 applied and whether a change to the plant Technical Specifications was involved. Maintenance of records and reporting requirements were also include )
However, the team noted that the procedural guidance was not consistent with current l NRC interpretations of the regulation discussed in Part 9900,"10 CFR Guidance," )
and NUREG-1606, " Proposed Regulatory Guidance Related to implementation of '
10 CFR 50.59". j Procedure SO123-XV-44 appeared to limit the scope of 10 CFR 50.59 to changes that l could impact the safety analysis of the plant. The procedure also stated that changes to secondary components shown on Updated Final Safety Analysis Report piping and ,
instrument drawings do not require a safety evaluation if there is no impact to the safety I of operatio CFR 50.59 permits the licensee to make changes in the facility as described in the safety analysis report unless the proposed change involves a change in the Technical Specifications or an unreviewed safety question. Although the statements in Procedure C,0123-XV-44 are broad and not well defined, the team noted that both secondary components and items described in the Updated Final Safety Analysis l I
Report may involve unreviewed safety questions regardless of whether they are directly called out in the safety analysis or have an obvious impact to the safety of operatio Additionally, the following guidelines provided in the instructions for Form SCE 26-548 i
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were not consistent with 10 CFR 50.59, as determined by the tea * Guideline 2 states that 10 CFR 50.59 includes certain changes to nonsafety-related systems, if the change can affect a safety-related syste ;
l The team noted that nonsafety-related systems could involve an unreviewed 1 safety question regardless of any direct effect on a safety-related syste * The guideline for Question 6 states that a new failure mechanism is not necessarily a new type of malfunction unless it has an effect not previously evaluated or bounded in the safety analysis repor The team noted that 10 CFR 50.59(a)(2)(ii) does not restrict the definition of a different type of malfunction as this guideline di * Guideline 6 states that an increase in consequences must involve an increase in dose to the public above the licensing limit. However, a note states, ". . . it a proposed change results in dose consequences which exceed Updated Final Safety Analysis Report values, but still remain below the licensing limit, identify this to management so that a decision can be made at appropriate management i levels."
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The team noted that 10 CFR 50.59 does not define an increase as only being a change that increases consequences above the licensing limit. On the contrary, 10 CFR 50.59(a)(2) defines an unreviewed safety question to include changes
-. where the consequences of an accident might be increase *
Instructions B.o states that where a change in the probability of occurrence or consequences of an accident / malfunction or margin of safety is so small or the uncertainties in determining whether a change has occurred is such that it cannot be concluded reasonably that probability / consequences / margin actually has changed, the change need not be considered an increase in the probability / consequences or a reduction in margin of safet The team noted that 10 CFR 50.59 (a)(2) states that a change involves unreviewed safety question if the probability of cccurrence or consequences of an accident or malfunction may be increased. Not only is there no permissive for smallincreases, but an unreviewed safety question exists if the probability might be increased by the chang Despite these inconsistencies between the procedural guidance and the current NRC interpretations of the regulation, the team noted that the licensee had management controls in place to ensure that changes involving unreviewed safety questions were not installed in the plant. Licensee personnel stated that they intentionally prepared procedures utilizing the guidance contained in NEl 96-07. This guidance conflicts with NRC interpretations as documented above. However, the team noted that the procedural guidance and training included management hold points for further evaluation when changes were proposed that resulted'in evaluations in the conflict areas. For example, a note following Guideline 6 in the instructions for Form SCE 26-548 stated, "If a proposed change results in dose consequences which exceed Updated Final Safety Analysis Report values, but still remain below the licensing limit, identify this to management so that a decision can be made at appropriate management levels."
To ensure that these management controls were effective, the team reviewed a number of recent modifications and changes. The team reviewed safety evaluations and evaluation screenings produced regarding changes made to the component cooling water and other systems. Additional evaluations were reviewed and evaluated as documented in Sections E1.2.1 and E8.1 of this inspection report. The team did not identify any cases where the inconsistencies between licensee procedural guidance and the current NRC interpretations of the regulation resuld in inadequate 10 CFR 50.59 evaluations. Records requested were readily retrieva% ar'd supported the subject changes. With the exception of the design control e - iocumented in Section E1.2.1, the safety evaluations reviewed were well documenteu The safety evaluations properly concluded that no unreviewed safety questions existed, indicating that the management controls were adequat .
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. Conclusions Procedural guidance related to the implementation of the requirements of 10 CFR 50.59 was not consistent with current NRC interpretations of the regulation. However, management controls, established in the procedure, required additional reviews should the consequences increase above Updated Final Safety Analysis Report limit Additionally, with the exception of the design control error documented in Section E1.2.1, the safety evaluations reviewed were well documented and properly conclud that no unreviewed safety questions existed indicating that the management controls were adequat E6 Engineering Organization and Administration (37550) !
l E6.1. Comouter Compliance for the Year 2000 l Inspection Scope i l
The team reviewed the licensee's program to evaluate computer systems and microprocessors for continued functionality beyond the year 2000. The scope, problems identified, and status were reviewed. Four computer systems were selected for further assessment: the digital radiation monitoring system, the core operating limit supervisory system backup computer, and the installed personnel contamination monitors, Observations and Findinas The team reviewed the " SONGS Y2K Weekly Progress Report," dated October 16,1998. The identification of all process components / applications with the potential to be affected by the change of millennium had been completed. This review resulted in 379 items or item types to be verified. To date,210 of the 379 detailed assessments required had been completed. The project was on schedule for completion in June 1999. The implementation of completion of most assessments were scheduled before January 1,1999, with final testing to be completed during the first half of the yea Significant line items in the licensee's plan were reviewed. The core operating limit supervisory system backup computer system was determined to have noncompliant software, because dates in the year 2000 and beyond could not be manually set. A software upgrade to remedy the problems is in progress. The RM-80 radiation monitoring system failed to restore to year 2000 after the system lost power and failed to automatically rollover at the end of the century. The licensee has decided to implement administrative controls to manually reset the year when needed. No hardware nor software upgrades are planned. Finally, the installed personnel contamination monitors had multiple year 2000 test failures. Licensee reviews indicated that all issues could be accepted as-is, or corrected by procedures for a manual input of the appropriate dat .
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A Conclusions The team determined that the licensee had established a reasonable program to identify and address computer components /cpplications that would not operate properly beyond the millennium. Corrective actions for the specific items revieweo were appropriate and being implemented in a timely manner. The schedule and schedule adherence supported completion of the project prior to December 31,199 E6.2. Enaineerina Backloa Insoection Scope The team reviewed Procedure SO123-XX-1," Action Request / Maintenance Order initiation and Processing," Revision 10. The team reviewed the licensee's engineering action request backlog and the manner in which the backlog was trended and tracke in addition, the team discussed the backlog with appropriate licensee personne Selected action requests reviewed are discussed in detailin Section E7.1 of this repor Observations and Findinas The action request process was used by the licensee to provide a single process for documenting the evaluation and resolution of all problems, concerns, activities, and conditions that could adversely affect or have the potential to adversely affect the safe operation of the plan From the trending information, the team found that in September of 1996, there were 996 open action requests, and in September of 1998, there were 2836 open action requests. The licensee acknowledged the adverse trend for action request closure, and explained that one of the reasons for the increase was that there was no minimum threshold for writing action requests. In ad2 ion, the licensee stated that each time an action request was written for an issue, each action to correct the issue was given an action request number and tracked separatel l To correct the adverse trend in action requests, the licensee stated that they were in process of implementing a work management tracking system. The work management tracking system would contain the corrective actions for an issue that were currently in the action request process. This would leave only the issue itself as an action reques j The action request could not be closed until all of the corrective actions were complete The licensee estimated that the work management system would be in effect by the beginning of 199 In addition, the licensee recently screened tho action request "cacklog and prioritized the items. Action requests, reviewed by the team during the course of this inspection were appropriately prioritize Conclusions The licensee had recognized an adverse trend with respect to the action request backlog, and was developing work management controls to address this issue. All-20-
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' outstanding action requests had recently been reprioritized. The action requests that were reviewed by the team during the course of the inspection were appropriately prioritized.
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E7 Quality Assurance in Engineering Activities (37550)
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E Action Reauests Inspection Scope
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The team reviewed selected action requests associated with the component cooling l
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water system. The team evaluated the action requests for adequacy of problem definition; problem resolution and timeliness; root cause consideration; and i consideration of extent of condition. The team discussed the action request process and some of the action requests with appropriate licensee personne ]
- Findinas and Observations
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Component Cooling Water Non-critical Loop Return Valve Closed Significantly
- - Faster than the Non-critical Loop Supply Valve j i
< The team reviewed Action Request 971001392, dated October 17,1997, which
- identified an unanalyzed condition in which one of the component cooling water non- ,
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. critical loop 28 inch air-operated butterfly return valves closed significantly faster than the fastest non-critical loop 28 inch air-operated butterfly supply valve. During a loss-of-coolant accident, the component cooling water non-critical loop inside i containment could be impacted by a high energy line break. A containment isolation !
actuation signal due to a high energy line break occurs when containment pressure
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reaches the high pressure setpoint of 5 psig. In response to the signal, the 28 inch ;
noncritical loop isolation valves begin to close. The team noted that the Technical Specification limit on closure time for the valves was 20.9 seconds after receipt of a containment isolation actuation signal. In addition, the licensee determined that the maximum differential closing time between the supply and return valves was two seconds to avoid excessive pipe support loading and water hammer. The valves were air-operated butterfiy valves with limit switches and control room indication for closur ,
During the last ASME Section XI inservice test, the Unit 3, Train A, Return Valve 3HV6218 closed in 4.8 seconds and the Supply Valve 3HV6212 closed in10.6 second The licensee determined these closing times using control room indication and not local indication. The licensee performed a test on September 16,1998, to investigate why the inservice test closure time for Valve 3HV6218 was significantly shorter than the
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closure time for the other valves. The test results indicated closure times of 16.34 :
seconds for Valve 3HV6218 and 17 09 seconds for Valve 3HV6212 using local indication. The licensee also found that the closure times were 10.0 and 15.69 seconds using control room indication. The licensee concluded that the closure times for the valves were within the two second limit and, therefore, all trains of component cooling water were operable. The licensee also determined that the control room indication for Valve 3HV6218 occurred early due to limit switch adjustment problems. The licensee-21-
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issued Action Request 980901120, dated September 18,1998, which documented the need for the maintenance department to reset the limit switch on the valv Action Request 980901120 also stated there was a need to evaluate the difference in the stroke time between the inservice test and the special test that was performed in September 1998. The inservice testing valve stroke times were much faster than the stroke times determined during the special test. Since the inservice test was conducted with no flow present and the special test was conducted under dynamic conditions, one possible cause of the difference in stroke times was the effect of fluid through the pip The licensee stated that they would perform additional tests or analysis during the next outage to determine the cause for the increased closing time under dynamic test conditions. To evaluate the risk associated with delaying actions until the refueling outage, the licensee reviewed their safety-related air-operated valves and assumed a similar increase in stroke time to account for possible dynamic effects. Using this conservative assumption, the licensee found only three groups of valves that might fall outside of their maximum opening or closing time. The licensee noted that these valves included a spring assist in the close direction (the direction of the safety function), and determined these valves would meet their closing times under dynamic condition The NRC plans additional review of the licensee's determination of the cause of the difference in stroke time during the static and dynamic testing of safety-related air-operated valves. This review will be tracked as inspection Followup item 50-361/9814-04; 50-362/9814-0 Non-Bellows Style Containment Penetration Relief Valves During review of Action Request 970101423, dated January 23,1997, the team noted that the penetration relief valves were non-bellows style, which would cause the setpoint to be susceptible to back pressure. If back pressure was present, it would be added to
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the spring force which would cause the valve to lift at a higher pressure than the design setpoint. The team noted that, during a design basis accident, up to 50 psig of back pressure would be imposed on the discharge of the relief valves which would cause the set pressure of the valves to increase from 150 psig to 200 psig. The system design pressure for the penetraticns was 150 psi For the CCW peaetration thermal relief valves, the licensee defined the worst case normal operating condition as a significant rise in temperature in the section of piping between the isolation valves at worst case normal containment pressures. They determined that it was not necessary to consider temperature conditions during a loss-of-coolant event or a main steam line break event, because these events are determined to be faulted plant events. In consideration of the faulted plant events, the licensee stated that they calculated the allowable internal pipe pressure for the piping system by using the ASME Section lli Code,1974 Edition Subsection NC-3641 equation. The licensee determined that the allowable internal pressure was greater than four times the relief valve setpoint pressure. As a result, the licensee concluded that containment temperatures following the design basis accident on these relief valves posed no significant impact on maintaining the structural integrity of the piping systems protected by the relief valve '
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The licensee stated that the relief valve setpoint should remain at 150 psig and the j valves should be tested for setpoints with no back pressure. The licensee concluded ;
this since the expected back pressure occurred only during faulted conditions such as a 1 loss of coolant accident or main steam line break inside containment. The licensee l
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concluded that resetting the relief valve setpoint to include back pressure would !
prematurely lift and relieve pressure along with loss of inventory during normal system operation which could lead to an unsafe plant conditio The team discussed this issue with the NRC program office. While a non-bellows style
! valve was not an optimal design 'or this application, the team concluded the design met l
the ASME Section lll requirement ;
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Component Cooling Water System Miniflow Valve Action Requests Action Requests 960400461,960900782,970501704, and 981000510 all involved problems with the Unit 3 CCW pump 024 miniflow control valve. Except for Action Request 981000510, the problems characterized in the action requests did not effect the operability of the valves. Action Request 981000510 considered Train A CCW as inoperable because of low accumulator pressure to the valve actuator, until the valve was electrically failed closed by de-energizing the valve operator control power. While not affecting operability, the team noted that Action Requests 960400481 and 960900782 had been unresolved since 1996, which did not appear to be timel Component Cooling Water Surge Tank Relief Valve Set Pressure Change to increase Tank Pressure Operating Range The team reviewed Non-Conformance Report 970800663, Revis5n 1, which discu.= cad revising the setpoints of the component cooling water surge tank relief valve. The setpoints of the two relief valves were increased to expand the allowable pressure operating range for the surge tank. Prior to the setpoint revision, filling the surge tt - I with makeup water resulted in a pressure increase in the surge tank above the setpo..,t of the relief valve. This resulted in loss of nitrogen from the tank that was not accounted for in the original nitrogen tank sizing calculation. The setpoints of the relief valves were increased to avoid opening the valves and releasing nitrogen, when makeup water was added to the surge tan The team noted that previous calculations determined the maximum setpoint of the relief valves assuming the component cooling water pump was deadheaded. The team reviewed Calculation M-0027-017, " Backup Nitrogen Supply for the Component Cooling Water Surge Tank," dated October 16,1997, which was performed to support the increase in relief valve setpoint. The team noted that in ordar to increase the setpoint of the relief valves, the licensee had to allow a mininnm flow of 5000 gpm through the pump, which allowed the maximum pressure in the surge tank to increase and still maintain the 150 psig design pressure at the pump discharg The 1974 Edition of ASME Section Ill, subsection ND-7411, required that the pressure rise no more than 10 percent above design pressure under any anticipated transien The team considered that requiring a minimum flow of 5000 gpm through the pumps-23-
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was'less conservative than the original calculation assumption that the pumps were t
. deadheaded, s
y The licensee further reviewed the piping between the pump and the discharge isolation V valve and noted that the installed piping was of a greater maximum wall thickness than ' 1
. assumed in the original system calculations. The licensee determined that if the failure l
of the nitrogen regulator occurred concurrent with the CCW pump discharge valve being i m
- inadvertently closed, the resulting pressure would still be within 1.1 times the design limit -
of the affected pipe segmen The team concluded the as-found design was acceptabl , 4 Conclusions Based on a review'of selected action requests, the team concluded that the action L request process was acceptably implemente i lE89 Miscellaneous Engineering issues (92903) :
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E8.1 1 Closed) Licensee Event Reoort 50-361/96-012: Condensate Storage Tank Outside its Design Basis 2 Backuround On February 26,1996, the licensee completed an evaluation of potential losses from Condensate Storage Tank T-120 following a seismic event. Licensee personnel had -
J- . determined that during the initial licensing process, the potential effects'of certain
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interactions with the piping and components downstream of the Seismic Category I
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isolation valves had not been properly reviewed. The results of the evaluation indicated that the 280,000 gallon minimum volume required by Technical Specification 3.7.6 did not provide reasonable assurance that 200,000 gallons would be available from
' Tank T-120 following a design basis earthquak i Inspection Followuo j l
The team found that Condensate Storage Tank T-121 provided a safety-related Seismic 1 Category I source of condensate to the auxiliary feedwater system. Condensate
- Storage Tank T-120 provided a nonsafety-related Seismic Category ll supply for normal condensate makeup to the hotwell for secondary system inventory control. In addition, T-120 provided an alternate source of water to Tank T-121 via a manually aligned ,
.' equalization line. Tank T-120 was surrounded by a Seismic Category I enclosure wal The enclosure was connected to TankT-121 via a gravity feed crosstie line and required local manual operator action to provide water from the enclosure to Tank T-121 in the event of a failure of Tank T-12 ,
- The licensee's compliance with 10 CFR 50, Appendix B, General Design Criteria 19 was
' assessed during initial licensing utilizing the guidelines in Branch Technical Position
- RSB 5-1," Design Requirements of the Residual Heat Removal System," Revision i NRC staff review had concluded that although San Onofre was not in full compliance
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with the branch technical position, Tank T-121 provided the primary shutdown capability and T-120 provided an " adequate alternate Seismic Category I" source of auxiliary feedwate The Safety Evaluation Report related to the operation of San Onofre Nuclear Generating Station Units 2 and 3 was documented in NUREG-0712. Section 9.2.4, " Condensate Storage and Transfer System," docunanted that a total of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of assured water supply would be acceptable. "This requirement necessitated the use of approximately 200,000 gallons of the condensate contained in the Seismic Category 11 tank."
At original licensing the design basis requiremeni to meet the Class 2 guidelines in l Branch Technical Position RSB 5-1, was not properly translated into the Technical l Specification 3.7.6 minimum volume requirements of 280,000 gallons. In February 1996, the licensee administratively increased the minimum level in Tank T-120 to 8 percent (382,207 gallons). Abnormal operating procedures were revised to ensure that I the secondary systems would be manually isolated from the system within 30 minute The team reviewed the level history available for Tank T-120. The tank was routinely maintained above 383,000 gallons as part of normal operating considerations. With the exception of the first operating cycle, reductions in inventory below this were related to outages or major power manipulations. Therefore, altnough an exhaustive review was not conducted, no examples of the licensee operating the plant with Tank T-120 at less than the administrative minimum volume for longer than the Technical Specification allowed outage time were identifie The team reviewed Calculation M-0050-017,"BTP RSB 5-1 Condensate inventory."
The assumptions and worst case analysis were reasonabic considering the design basis of Tank T-120. The calculation determined that 330,351 gallons were required in the tank during worst case conditions to ensure that 200,000 gallons would be available for plant cooldown, if required. Based on this calculation, the licensee is currently in the process of applying for a Technical Specification amendment to increase the required volume in Tank T-120 from 280,000 gallons to 360,000 gallon The required volume in Tank T-121 was 144,000 gallons combined with a calculated requirement of 330,351 gallons in Tank T-120, the total condensate volume required for compliance with the branch technical position would have been 474,351 gallons. Had the licensee only maintained the Technical Specification minimum volumes in the tank (280,000 gallons for Tank T-120 and 144,000 gallons for TanP T-121), the condensate volume maintained in the tanks would have been 424,000 gallons. This would have represented a maximum deviation of 10.6% frnm the total condensate volume required for compliance with the branch technical position. Therefore, the safety significance was determined to be lo The team walked down plant areas associated with Abnormal Operating instruction SO23-13-3," Earthquake," to ensure that isolation valves were accessible and that operators were capable of isolating the tank within 30 minutes. Isolating the tank involved isolating one valve in each of two flow paths. The team determined that operators could have isolated the larger flow path by closing Valve MUO92, within 30 minutes. Closing Valve MUO92 isolated the major contributor to postulated losses from Tank T-120 or the associated enclosure. The team questioned the ability of operators-25-
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to access and close Valve HV5715 that isolated the second flow path, within 30 minutes, because it was located in a potentially life-threatening environment. Licensee management stated that operators would be able to access the valve within 30 minute Furthermore, additional time to access the confined space was available once Valve
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MUO92 was close In summary, the licensee determined that Technical Specification 3.7.6 did not provide reasonable assurance that 200,000 gallons would be available from Tank T-120 following a design basis earthquake; however, no clear examples were identified in which the tank volumes had been maintained below the allowable volumes.
! Additionally, the maximum possible deviation between the water supply required to comply with the Branch technical position and the required minimum Technical
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Specification inventories were approximately 10.65 Therefore, the safety significance ]
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was determined to be low. Licensee actions following identification were sufficient to ;
correct the condition and ensure compliance with the Branch Technical !
Position RSB 5-1. This failure to assure that applicable design basis parameters were I translated into the Technical Specifications and associated surveillance test procedures was in violation of 10 CFR 50, Appendix B, Criterion 111," Design Control." This nonrepetitive, licensee-identified and corrected violation is being treated as a noncited ,
violation consistent with Section Vll.B.1 of the NRC Enforcement Policy j (50-361/9814-05;50-362/9814-05). ,
l E8.2 (Closed) Licensee Event Report 50-361/98-009-01: Condensate Storage Tank Outside Design Basis Due to Procedural Error ;
1 Backaround On April 28,1998, an engineer reviewing calculations of potentiallosses of water from Tank T-120 discovered that the steps in Abnormal Operating Instruction SO23-13-3,
" Earthquake," Revision 5, to isolate the tank following an operating basis earthquake had been inadvertently deleted. The administrative requirement to maintain the level in Tank T-120 greater than 382,207 gallons remained in place. However, operator action following a design basis earthquake was still necessary to assure that the design basis 200,000 gallons water volume was maintained. The steps had been inadvertently deleted in August 1996 as a result of a word processing error, Inspection Followuo As corrective action, the steps were replaced on April 29,1998. The event was reviewed with the individual involved and an additional barrier for surveillance procedure, operating procedure, and abnormal operating procedure revisions were establishe The team determined that the design basis requirement to meet the Class 2 guidelines in Branch Technical Position RSB 5-1 were not properly translated into operating procedures following the revision of Abnormal Operating instruction SO23-13-3 that inadvertently removed the step As documented in Section E8.1 of this inspection report, the water inventory maintained in Tank T-121 was considered sufficient to achieve safe shutdown. Additionally, had a seismic event caused Tank T-120 to drain, water inventory in Tank T-121 would permit-26-
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approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to isolate and refill Tank T-120 from one of the other alternate
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water sources shown on Figure 10.1-7 of the Updated Final Safety Analysis Repor Additionally, astute operators could have identified a decreasing tank level and isolated Tank T-120 preserving some tank inventory. The team concluded that although the
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procedures were inadequate to ensure compliance with the design basis requirements-in Branch Technical Position RSB 5-1, the auxiliary feedwater system remained ;
operable and capable of performing its intended safety function.
j The team determined that this failure was not a repetition of the noncited violation discussed in Section E8.1, because the cause of the omission was related to a word
. processor error. i his failure to assure that applicable design basis parameters were l translated into operating procedures was a violation of 10 CFR Part 50, Appendix B, Criterion Ill," Design Control." This nonrepetitive, licensee-identified and corrected
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violation is being treated as a noncited violation, consistent with Sec1on Vll.B.1 of the NRC Enforcement Policy (50-361/9814-06; 50-362/9814-06).
E8.3 (Closed) Licensee Event Report 50-362/98-012-00 and -01: Component Cooling Water Relief Valve Inoperable - Wrong Set Spring Installed Due to Vendor Data Error
' Backaround ,
The licensee identified that an incorrect spring and lift stop were installed in Train A .
Component Cooling Water Surge Tank Relief Valve 3PSV-6358. The incorrect valve l components could have prevented the Train A component cooling water system from venting during a CCW over-pressure transient caused by a postulated seismic event, which potentially placed the plant outside its design basis. The licensee performed an investigation of the circumstances that led to these failures and concluded that Unit 3 i Train A CCW was inoperable between March 2,1987 and February 21,1992. The licensee also determined that the incorrect spring was installed as a result of a vendor error, and that the lift stop was not cut to the correct length, because the vendor had not included installation instructions with the lift stop, Inspection Followuo The team reviewed LER 50-362/98-012-00 and -01 and the action requests associated with the incorrect spring installation and the incorrect lift stop length, incorrect Component Cooling Water Surge Tank Relief Valve Set Pressure Because of Incorrect Spring installation The team reviewed Action Request 97100147, dated October 28,1997. The action request identified that, between March 1987 and February 1992, the set pressure of Relief Valve 3PSV6356, which was mstalled on the component cooling water surge tank to prevent overpressure, had an incorrect spring installed. As a result, the relief valve had a setpoint of 150 psig instead of the required 45.5 psig. This valve was the only relief valve installed on the surge tank at the tim .
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Licensee Event Report 1998-012-01, dated July 1,1998, stated that the incorrect spring in the relief valve could have prevented the Train A component cooling water system from venting during an overpressure transient caused by a seismic event. The licensee '
determined that additional nitrogen pressure from a failed pressure regulator would cause the system to exceed its 150 psig design pressure when combined with the maximum system operating pressure. However, based on a review of the available design margins the licensee stated that the train was available and able to perform its intended safety function. The licensee stated that the pressure regulator valve for the surge tank was not seismically qualified during this time period. However, the same make and model were seismically qualified for other applications at the plant. The licensee concluded that the Train A of component cooling water system for Unit 3 was inoperable between March 1987 and February 1992. Since Unit 3 Train A of component cooling water was inoperable for more than its Technical Specification allowed outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the licensee identified that they had been in violation of Technical Specification 3/4.7.3. This non-repetitive, licensee-identified and corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the NRC dnforcement Policy (50-362/9814-07).
Incorrect Component Cooling Water Surge Tank Relief Valve Set Pressure Because of Incorrect Lift Stop Installation The team noted that Action Request 97100487 also identified that, between December 1993 and April 1997, an incorrect lift stop was installed in the surge tank Relief Valve 3PSV6356, which resulted in gagging the valve shut and making it inoperabl The licensee stated that, in 1992, the surge tank design was upgraded to include a safety-related seismically qualified backup system. During this upgrade, a second safety relief valve, non ASME Section lil, was installed on the surge tank to protect the backup nitrogen system piping its case the pressure regulator valve failed and overpressurized the system. The licensee stated that this valve also protected the surge tank and concluded that, even with Valve 3PSV6356 inoperable, the component cooling water was still operable during this tim The team noted that a block valve was installed between the surge tank and the new relief valve. The team determined that in accordance with ASME Section ill,1974 Edition subsection ND-7153, a block valve was not allowed to be installed between a relief valve and the system it was protecting unless the block valve was installed with positive controls and interlocks to ensure that rekving capacity requirements were me The team noted that the block valve was an open manual valve arsd was not locked open. However, the licensee stated that it was difficult to reach and it was continually open during the period of concern. While the installed configuration did not comply with ASME Section Ill, the team agreed that second relief valve was functionally equivalent and the component cooling water system was still operable. The iicensee properly identified and corrected this violation of 10 CFR 50.55a, which requires compliance to the ASME Boiler and Pressure Vessel Code. This non-repetitive, licensee-identified and corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy (50-362/9814-08).
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E (Closed) Licensee Event Report 50-361/98-014: Emergency Feedwater Actuation Signal Outside Design Basis Backaround l On August 12,1998, operators in simulator training noticed that the failure of Vital Bus 2 I unexpectedly resulted in the steam-driven auxiliary feedwater pump starting and discharging to Steam Generator E-088. A subsequent evaluation revealed a design flaw in the circuit as installed in the plant. The licensee determined that during a postulated main feedwater line break, a failure of Vital Bus 2 would result in auxiliary ,
feedwater spilling from the broken feedwater line. This represented a more limiting accident case than previously evaluated for the Condensate Storage Tank T-121 capacit Inspection Followuo I The team reviewed controlled drawings related to this concern and determined that the failure of Vital Bus 2 would have caused the postulated event. The licensee concluded that operators would manually isolate the diverted auxiliary feedwater flow within 30 minutes. Considering main control room instrumentation, annunciation, and emergency response procedures, this appeared to be reasonable. The team reviewed Emergency Operating Instruction S023-12-5, " Excess Steam Demand Event," Revision 15 and determined that operator actions were reasonable and would appropriately i address the loss of condensate through a broken main feedwater line in a timely manner. This response would limit the loss of condensate to approximately 30,000 gallons. As discussed in Section E8.1 of this inspection report, an adequate alternate source of auxiliary feedwater supply has been maintained in Condensate Storage Tank T-120. The Technical Specifications provide for a minimum of 280,000 gallons to be maintained in that tank. The team concluded that sufficient auxiliary feedwater was available until the design flaw could be corrected. The licensee noted that this condition has no actual safety significance because additional condensate is available for shutdown, and postulated transients are bounded by Updated Final Safety Analysis Report analyses. The licensee also stated that the two previously reported conditions of the condensate storage tank being outside of its design basis are unrelated. This failure to evaluate the most limiting condensate storage requirements for response to a main feedwater line break constitutes a violation of minor significance and is not subject to formal enforcement actio V. Management Meetings X1 Exit Meeting Summary The team met with licensee representatives on November 6,1998, to conduct an exit intervie The licensee acknowledged the team's findings. During this meeting, the team leader noted that team personnel had not reviewed proprietary documentation during the course of the inspectio .. . .. . .. .. . - .. . . . - .
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ATTACHMENT
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SUPPLEMENTAL INFORMATION -
PARTIAL LIST OF PERSONS CONTACTED Licensee D. Axline, Engineer G. Gibson, Manager Compliance R. Krieger, VP Nuclear Generation J. McGaw, Manager 10 CFR 50.59 Program D. Niebruegge, Manager Station Technical D. Nunn, VP Engineering and Technical Services R. St. Onge, Manager P. Scofield, Supervisor D. Van Buskirk, Engineer J. Stoessel, Engineer M. Wharton, Manager Engineering Design R. Waldo, Manager Operations NRC J. Russell, Resident inspector INSPECTION PROCEDURES USED
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'37550- Engineering 92903- Followup - Engineering 93809 . Safety System Engineering Inspection (SSEI)
37001 10 CFR 50.59 Evaluations j:
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o ITEMS OPENED. CLOSED. AND DISCUSSED Opened 50-361/9814-01; VIO c ailure to assure that appropriate quality standards were 50-362/9814-01 specified and included in design documents for the electrical control of the component cooling water miniflow control valves as required by 10 CFR 50, Appendix B, Criterion lli (Section E1.2.1).
50-361/9814-02; IFl Evaluate the licensee's method for assuring appropriate 50-362/9814-02 quality standards for the non-Class 1E components that had -
been evaluated as capable of withstanding a seismic event in Report M86420 (Section E1.2.1). ,
50-361/9814-03; URI The failure to update heat exchanger operability curves to 50-362/9814-03 address thermal performance data is unresolved pending NRC evaluation of the results of planned testing to better estimate the total tube wall fouling of these heat exchangers and NRC review of the licensee's evaluation of past operability (Section E2.2).
50-361/9814-04; IFl Evaluate the licensee's determination of the cause of the 50-362/9814-04 difference in stroke time during the static and dynamic testing of safety-related air-operated valves (Section E 7.1).
50-361/9814-05; NCV Failure to assure that applicable design basis parameters 50-362/9814-95 were translated into the Technical Specifications and associated surveillance test procedures was as required by 10 CFR 50, Appendix B, Criterion lil," Design Control,"
(Section E8.1).
50-361/9814-06; NCV This failure to assure that applicable design basis parameters 50-362/9814-06 were translated into operating procedures was in violation of 10 CFR 50, Appendix B, Criterion lil," Design Control,"
(Section E8.2).
50-362/9814-07 NCV Train A of component cooling water system for Unit 3 was inoperable between March 1987 and February 1992 in violation of Technical Specification 3/4.7.3 (Section E8.3).
50-362/9814-08 NCV Failure to assure compliance with the ASME Boiler and Pressure Vessel Code in violation of 10 CFR 50.55a (Section E8.3).
Closed 50-361/96-012 LER Condensate storage tank outside its design basis (Section E8.1).
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50-361/98-009-01 LER Condensate Storage Tank Outside Design Basis Due to Procedural Error (Section E8.2).
50-362/98-012-00- LER Component Cooling Water Relief Valve Inoperable - Wrong i and -01 Set Spring Installed Due to Vendor Data Error (Section E8.3). l
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50 361/98-014 LER Emergency Feedwater Actuation Signal Outside Design l Basis (Section E8.4).
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50-361/9814-05; NCV Failure to assure that applicable design basis parameters 50-362/9814-05 were translated into the Technical Specifications and associated surveillance test procedures was as required by i 10 CFR 50, Appendix B, Criterion lil," Design Control," l (Section E8.1).
50-361/9814-06; NCV. This failure to assure that applicable design basis parameters 50-362/9814-06 were translated into operating procedures was in violation of 10 CFR 50, Appendix B, Criterion Ill," Design Control,"
(Section E8.2). ;
50-362/9814-07 NCV Train A of component cooling water system for Unit 3 was i inoperable between March 1987 and February 1992 in violation of Technical Specification 3/4.7.3 (Section E8.3).
50-362/9814-08 NCV Failure to assure compliance with the ASME Boiler and Pressure Vessel Code in violation of 10 CFR 50.55a (Section i E8.3).
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.LCFRi Code of Federal Regulations -
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'CCN ;calculction change notice
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1 'EOP- .. emergency operating procedure
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Y EPRil Electric Po'wer Research l'nstitut l
-. GDC 1 general design criteria : 1
, i gpm ; gallons per minute ' -i IFl: - inspection followup item" !
"LER '
licensee event report iNPSHl- ' net positive suction head
. hsl - pounds per square inch
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W-$SEl safety system engineering inspection i
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TS LTechnical Specificatio JUFSAR- Updated Final Safety Analysis Report
T URl! unresolved item -l
US unreviewed safety question I
.Vdc ' volts direct current -
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DOCUMENTS REVIEWED
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SAFETY EVALUATIONS
REVISION
~ NUMBER DESCRIPTION j
..c - 40127A- ' FCN F11676M: Restore control power to the CCW
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November 30,1995
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. minimum flow valves (2HCV-6537,2HCV-6538, and ;
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' 2HCV-6539) to its original design configuration - !
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' F1447'1 M ': Installation of globe valves for throttling CCW flow ; 2 3
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' SAFETY EVALUATIONS ,
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NUMBER. DESCRIPTION REVISION ;
F7237JK - Increase setpoint of CCW surge tank nitrogen 0 F7346J . Delete normal makeup fill valve' automatic function 2
. NPF10/15f DRAFTincrease CST T-120 required water volume - Draft ACTION REQUESTS-NUMBER DESCRIPTION REVISION 940200022.' Relief valve was found incorrectly set 50 psi February 11,1994 over the required set pressure 1960100345 ~ The backpressure regulating valve is at a - January 11,1996-location where maintenance activities are difficult
.960101028- CCW cross train l$akage exceeds the limit ~ January 28 -1997
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specified in SO23-3-3.25 9604004811 - [3HCV-6537 - CCW P024 miniflow block - April 9,1996 valve) Air leak on back side of valve on -
- tubing fitting 960900782 [3HCV-6537 - CCW P024 miniflow block September 18,1996 valve) Valve leaks by seat and pointe ,
indication not exactly zero when valve fully l shut; when P025 aligned to Train A and running, P024 miniflow has noticable flow noise; need to have seat repaired or valve internals adjuste Approximately 150 design change October 4,1996 packages are in some phase of the turnover process 961201253 Valve seat failed set leak test December 17,1996 ;
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970101423' Relief valves protecting containment January 23,1997
. penetrations were not in the ASME Section
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l970401058- Breaker failed testing (Section 6.2.5 of April 18,1997 i SO23-1-2.52) i
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970401702 The low detta P function of the switch has April 30,1996 failed repeatedly in the past
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d ACTION REQUESTS NUMBER DESCRIPTION REVISION 970401970 Tha type and style of pressure control valve April 30,1997 for the equipment is incorrect 970401970 The type and style for the pressure control April 30,1997 valve for this equipment is incorrect 970500103 The DCP for undervoltage circuit fuse May 2,1997 failure alarm was never implemented 970500606 CCW heat exchangers have different May 9,1997 thickness gaskets than the drawing shows 970501704 (3HCV-6537 - CCW P024 miniflow block May 22,1997 valve] Valve indicates 15% open with no flow noise when 'A' Train CCW is operating; indication needs to be adjuste Identified that PFC 2/3-84-044 should be June 10,1997 canceled 970700717 Block heater found hot but water did not July 15,1997 appear to be circulating 970800069 Level in the Unit 3 train A is lowering at a August 2,1997 rate of about 3.5 % per day over the last few days 970800663 The method for adding makeup water to the September 30,1997 CCW surge (ank is inconsistent with the design basis of a backup nitrogen system 4 971001392 One (,f the CCW non-criticalloop return October 27,1997 valves closed significantly faster than the ,
fastest supply valves i 971001487 Whe 1 CCW relief valve was disassembled, October 28,1997 1 it waa found that the lift stop had never been cut to the proper length and the wrong spring was installed 971100831 During a review it was noted that there was November 14,1997 a conflict between the stated set pressure !
and back pressure i l
971200122 Need to review PFC 3-85-6436 to either December 3,1997 l cancel or fund it
i
!
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_ _ _
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, .. . _._ _ . . _ . . . . . .,- _ . _ . _ . _ _ _._.... .m _
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i 1 ACTION REQUESTS I NUMBER < DESCRIPTION REVISION :
971200523 Modification for backflow preventors was ' Decembe'r 10',1997 prepared and then canceled. The DCP l
- needs to be formally canceled I
-
971200535: Partial work on PFC 2/3-83-215 was December 10,1997. - l completed but has not been completed l 971200686 . information provided by vendor states that - December 12,'1997
,
' the installed valve shaft may not be correci 971200914- .
Review open paper work of PFC 2/3-88-012 . December 16,1997 to determine requirements to close it .
97120096 Review open paperwork of PFC 2/3-089- December 17,1997 044 to determine requirements to close it
-
- 971200984? One'or more documents are required to December 17,1997 support closure of DCP 2/3-6742 971200992 One or more documents are required to December 17,1997 i support closure of. DCP 2/3-6818 l
.971200996 . Review open paper work to determine December 17,1997 requirements to support closure of PFC 2/3-87-017:
980501402 CCW train A surge tank level continuously May 17,1998 lowers-980600241- :
Level indicators show that the Units 3 & 3 . June 2,1998 -
have ccw cross leakage -
-980801720 Calculation J-EGA-071 determined that - August 27,1998
, ';
<
installed pressure gauges PI-6416/6422 did '
not meet tolerance requirements for TS surveillance and recommended appropriate M&TE for nitrogen regulator calibratio "380901120 IST stroke time data for CCW non-critical September 18,1998 loop supply and return valves is different -
..
from stroke time determined under flow conditions '--
'
'981000510 Solenoid continually ports air when miniflow October 8,1998 L ' valve is' shut; solenoid valve may need to be L ;c replaced; miniflow valve is operating c correctly. (3HY-6537 - CCW P024 miniflow
.. block valve]
-
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ACTION REQUESTS
- NUMBER
'
DESCRIPTION REVISION 981100386 Function of CCW miniflow valves may not _ November 5,1998 l have been appropriately considered during !
preparation of Report M8642 !
981100405 . Documentation of deviation to IEEE Std 308 November 6,1998 for limiting degradation of 125 Vdc 1 distribution system 981100409- Justification for acceptance criteria November 6,1998 presented in Section 5.13 of Calculation
'
E4C-098 l
'
'98080637000- Equipment ID: S21203ME001, . . . November 3,1998 Drill / Tap Existing Plugs . . .
981001913--03 Action Request Assignment Report, Heat November 5,1998 Exchanger Performance Testing item
- 981001913--03 - Operability Assessment 981001913--05 Action Request, Heat Exchanger October 28,1998
'
Performance Testing, including item #
~
981001913-05 - Initial Operability '
Assessment 98110031000 Equipment ID: 026-44326, Drill / Tap November 3,1998 Approx. 40 Tube Plugs . . .
98110031000 Equipment ID: 026-44326, Drill / Tap November 3,1998 Approx. 40 Tube Plugs . . . adding Revised -
Engineering Evaluation / 50.59 Evaluation NONCONFORMANCE REPORTS NUMBER DESCRIPTION REVISION
'
930600045 [2HY6538 - Air supply solenoid for HCV-6538] Control 0
) circuits are nether safety related nor separated from those of redundant valves 970601693 Installation of a jumper wire to the thermocouple terminals 0 and disconnecting the heater field leads 970800799- Disconnection of field wiring of CET L-16 and installation of a 0 jumper at the panel 971000764 Install a jumper wire to the thermocouple terminals 0-8-
.
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NONCONFORMANCE REPORTS
>
NUMBER- DESCRIPTION REVISION I
. 971100591 Remove an inoperable.HJTC level from service by installing 0 ;
1 jumper wires ' I
980201481 Removed phone booths, fire hose and extinguisher cabinets 2
. since they did not have qualified coatings
^
980300506 Removed phone booths, fire hose and extinguisher cabinets 1 i l since they did not have qualified coatings I 980301693 . Pressurizer heater was electrically disconnected 2 1
,
-
PROCEDURES NUMBER DESCRIPTION REVISION ,
JS-123-103C Setpoint/LoopAccuracy Calculation Methodology 2 i SCE 26-548 Engineering Design Program 10CFR50.59 Safety 3 ,
Evaluation SO123-V-5.10 Temporary Facility Modification 6 SO123-XV-44 - Guidelines for Determining when a 10 CFR 50.59 Safety 2 Evaluation is Required SO123-XV-5 Nonconforming Material, Parts, or Components 9 SO123-XV- Temporary Modific' ation Contro! 2 SO123-XVI-1 Confined Spaces 4
. SO123-XX-1 1 SS2 Action Request / maintenance order initiation and 10 processing SO123-XXX- Control of Licensing Document Changes 2 SO2-II-11.1 A Surveillance Requirement Unit 2 ESF Train A Loss of 1 Voltage (LOVS), Degraded Voltage (SDVS, DGVSS),
and Sequencing Relays and Circuits Test
' SO23-12-5 ' Excess Steam Demand Event 15 SO23-12-6 Loss of Feedwater 16 SO23-13-3 Natural Disaster / Severe Weather . 1 SO23-13-3 ' Post Operating Basis Earthquake Inspections ' 5 SO23-15-5 ' Annunciator Panel 53B, MFP-K0OS/ Condensate 3
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PROCEDURES-NUMBE DESCRIPTION . REVISION -
'
SO23-2-17 ~ Operating instruction - Component Cooling Water J .12
,
- System Operation
. SO23-3-3.13 Suiveilance Operating instruction - Containment Cooling _ 7 x - / Spray Monthly Tests c ,
i
'
q J SO23-3-3.18 . Operator Surveilance Test - Component Cooling / 9'
- Saltwater System Tests ,
LSO23-3-3.18 _ Operator Surveillance Test - Component Cooling /. 9 Saltwater System Tests SO23-3-3.25 Technical Specification Surveillance Log ~ 7
'
, ,
SO23-3-3.2 Weekly Electrical Bus Surveillance 11
'
SO23-3-3.3 Surveillance Operating Instruction - Component Cooling 6
'
- Water System Quarterly Valve Test
-
1S O23-3-3.3 Component Cooling Water System Quarterly Valves Test -
SO23-3-3.3 Surveillance Operating Instruction - Component Cooling ~6 Water System Quarterly Valve Test ;
SO23-3-3.3 Surveillance Operating Instruction _- Component Cooling 4 Water Valve Testing - Cold Shutdown and Refueling Interval SO23-3-3.60.3 - Component Cooling Water and Seismic Makeup Pump - 1 Test
^
SO23-6-15 - Operation of 125 VDC Systems 11 SO23-6-33 Ground Isolation 1 SO23-1-2.52 ' Containment Penetration Circuit Breaker Overcurrent 10
- Test SO23-1-8.235 Cold Bench Testing and Calibration of IST Program 1
- Safety Relief Valves j - SO23-1-8.88 ' Cold Bench Testing and Calibration of Non-IST Program 4 ASME 111, ASME Vill, and Non-ASME safety / relief Valves SO23-il-9.14 : Electronic Differential Pressure and Pressure Transmitter 1 Calibration
' SO23-V-3.13 Engineering Procedure -' Component Cooling Water 5 (CCW) Train Leakage Monitoring-10- )
!
.
, _ _ . . .- . .. . . . . . _ _ _ .
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PROCEDURES NUMBER- DESCRIPTION REVISION ,
i SO23-V-3.25 Engineering Procedure - Component Cooling Water Heat 3 J Exchanger Testing
'
SO23-V-3.2 Engineering Procedure - Component Cooling Water Heat 0 Exchanger Hydraulic Testing l SO23 V- Engineering Procedure -Inservice Testing of Pumps 9 Program ]
CALCULATIONS NUMBER DESCRIPTION REVISION A-94-E-00 CCW Miniflow Valves Circuit Analysis 0
!
.c E4C-014 Diesel Generator Sizing 6 j E4C-016 ESF Sequencing 5-
.
E4C-017 125 Volt Battery DC System Sizing 15 E40-082 System Dynamic Voltages During Design Basis 1 .
- Accident l
'
E4C-084 Unit 2 MCC Control Circuit Voltage Analysis 0 I
E4C-088 ' Emergency Diesel Generator Loading 1 thru CCN-18 E4C-090 Auxiliary System Voltage Regulation 1 E4C-098 4 kV Switchgear Protective Relay Setting 1 Calculation thru CCN 21 E40-109 Class 1E 125 VDC System Protection Calculation 0 l thru CCN-5 EC-374 Engineering Calculation - CCW Heat Exchanger 1 Performance U2C8 RFO EC-399 Engineering Calculation - CCW Heat Exchanger 0 Pedormance U2C9 RFO
~ J-EG A-010 CCW Surge Tank Low-Low Setpoint 1
,
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H n : CALCULATIONS '
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NUMBER 1- ' DESCRIPTION : REVISIC N .-
-
'e' ;, ? J-EGA-017 -
'
J Total Loop Uncertainty for CCW flow from HXs . E001A/E002B (2FT-6243/6248) " .
, , + <
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Performance Measurement - i
. . . . . ,
~ (J-EGA-025 ' i ~ CCW Low Pressure Pump Autostart Setpoint - '
-,
f , 12(3)PCL-6362-2 & 2(3)PCL-6363-1; ;
c
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T J-EGA'-040 - Scaling Calculation for'CCW Surge Tank Levell O
~
. Indication
=3
-
J-EGA-04 l SAtpoint Calculation for CCW Surge. Tank Level ~
, Indication and Alarms - . thru CCN-4
,N '
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. J-HEA-010 Scaling Calculation for Primary Plant Makeup . '
0- ;
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Storage Tanks LevelL ,
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.-J-HEA-010 : Scaling Calculation fee the PPMST Level 0
!
.e , Indication
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J y: K ( J-HEA-011 Setpoint Calculation for PPMU Storage Tank - 1-LevelIndicator and Alarms 4 thru CCN-1- j ,
l EJ-JEA-0111 Setpoint Calculation for the PPMST Level 0 i: -
'
-Indication and Alarms b 1Various' Room Flooding Alarm Sensor Setpoint 1'
'
L J-RNA-015 g
'
Calculation L y ,
' M-0050-017 - ' BTP RSB S-1 Condensate Inventor ; M-026-003 ; CCW Surge Tank Pressure 2
,
.
1- ~ M-026-011 CCW Flow / Pressure Distribution Analysis 1, CCN 1
{l 1 M-027-012 Estimated CCW Pump NPSHA During Selected 0
'
- . Transients y ; M-027-017. - -. Backup Nitrogen Supply for the CCW Surge Tank 0, CCN-1,
'
'
ICCN F-244
- .' , i M-027-017 (3ackup Nitrogen Supply for theCCW Surge Tank
.
0, ICCN-3 E - M-027-017 Backup Nitrogen Supply for the CCW Surge Tank 0, CCN-1,
> ICCN F-244
.
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' ' CALCULATIONS
!
NUMBER DESCRIPTION REVISION
M-027-017 Backup Nitrogen Supply for the CCW Surge Tank 0
' M-027-020 - _ CCW Safety Related Makeup System Hydraulic . Rev. O, CCN-1,
.
' Calculations- ICCN F-580
,
M-027-020 ' CW Safety Related Makeup System Hydraulic C _
-0-Calculations
' M-027-020 CCW Safety Related Makeup System Hydraulic 0 Calculations
,
M-027-023 CCW Heat Exchanger Operability (Units 2 and 3, 0, ICCN-1
- based on Cycle 8 results)
CCW Heat Exchanger Operability (Unit 2 only -
~
-i ? - M 027-023~ O based on Cycle 8 results)-
M-DSC-235 Diesel Generator Load Verification - Mechanica! 0'
Equipment BHP Requirements - thru CCN-6
. N-4098-1 : Radiation Monitor System Parameters 5
'
N-720-3 Radiation Monitor Setpoint Conversions 2
-
ICCN C-1
,
None PSV 5403 / 5404 Additional Calculation provided November 5-6.1998 by Licensee 11/5-6/98
. None PSV'6356 / 6359 Additional Calculation provided November 4-5,1998 by Licensee 11/4-5/98 DESIGN CHANGES NUMBER- DESCRIPTION REVISION Pre lubrication modification of the Units 0 DCP2/3-6754.00SM 2 & 3 diesel generator tube oil system DCP2/3-6742.07SM Component cooling water safety related 0 make-up system DCP2/3-2077.00SE Containment ECUS Sequence Time O Modification DCP2/3-6818.00SM Modification of the diesel generator 0 starting air system-13-p ,
" ~5-C- __ . ._ - _
_ .
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- a:
DESIGN CHANGES-NUMBE DESCRIPTION ' REVISION
' DCP3 970.1J ' Revise CCW cross tie isolation valve 0 interlock controls FC.N F7237J Change Setpoint tolerance for Type 1 223/233 Valves
.
FCN F7346J . Remove control devices to prevent 3~
automatic actuation of the water fill valve FCN F14471M install valves to limit normal make up 26 flow t
FCN F10252E One Line Diagram, N-1E 125 V'dc 0 Distribution Panels 2DSP1,2D5P2, i ,
2D5P3,2D5P4
- PFC2/3-89-044 Installation of refueling cavity locallevei 0 indicator
- PFC2/3-88-012 Install chargers in battery rooms to allow 1 single cell charging of spare cells -
'
PFC2/3-85-038 Modification of hinge pins on Anchor 0 l Darling check valves in the AFW pump
,
steam system PFC3-85-6436 Provide ' vent / drain line to containment 0 spray system upstream of check valve PFC3-87-059 Replace carbon steel valve components 0 '
with boric acid corrosion resistant stainless steel parts
- TCN 3-2 Add Calculation M-0041-094 to DBD 3 SO23-780 TFM-98-BBA-001 Install a low resistance jumper on the 1 control terminals to eliminate a nuisance alarm TFM-93-GJA-001 Remova 3RE7823-2 (FHIS) auto start 0 frorr ' 335 emergency chiller i
n-14-
a O
DRAWINGS NUMBER DESCRIPTION REVISION 30107 Electrical One Line Diagram: 4160 V Switchgear Bus 12 2A04 30109 Electrical One Line Diagram: 4160 V Switchgear Bus 13 2A06 30118 Electrical One Line Diagram: 480 V Load Center 30137 Electrical One Line Diagram: 480 V MCC 2BE (ESF) 25 30142 Electrical One Line Diagram: 480 V MCC 2BJ (ESF) 22 30173 Electrical One Line Diagram: 125 Vdc Distribution 17 Switchboard Bus 2D1 30174 Electrical One Line Diagram: 125 Vdc Distribution 16 Switchboard Bus 2D2 30182 Electrical One Line Diagram: 120 Vac Vital Bus 2YO1 24 30183 Electrical One Line Diagram: 120 Vac Vital bus 2YO2 22 30186 Electrical One Line Diagram: 4160V Switchgear Key 1 Interlock Schemes 30587 Elementary Diagram: 2HV-6569 (CCW makeup pump 12 discharge to CCW Train A)
30588 Elementary Diagram: 2HV-6570 (CCW makeup pump 13 discharge to CCW Train A)
30690 Elementary Diagram: 2MP-1019 (CCW makeup pump 11 from PPMST)
30700 Elementary Diagram: CCW Pump 025 15 30701 Elementary Diagram: CCW Pump 024 17 30702 Elementary Diagram: CCW Pump 025 18 30704 Elementary Diagram: CCW Pump Suction HV-6222A 9 30705 Elementary Diagram: CCW Pump Suction HV-62228 5 30706 Elementary Diagram: CCW Pump Suction HV-6224A 8 30707 Elementary Diagram: CCW Pump Suction HV-6224B 6 30708 Elementary Diagram: CCW Pump Disch HV-6226A 5 30709 Elementary Diagram: CCW Pump Disch HV-6226B 4-15-
_ . _., . . . _ . . . _ _ . . _.. - _ . . - . . . . - - _ _ . .. . _ . . . _ _ _ _
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DRAWINGS : -
'
NUMBER-
'
q " DESCAIPTION : REVISION -
. 30710a Elementary Diagram: CCW Pump Disch HV-6228A 6- l
~
30711" 1,
. Elementary Diagram: CCW Pump Disch'HV-6228B' -- 7. .
o -
- 'I
- 30712'.' Elementary Diagram: CCW From Cont isol HV-6216 20
U. .. :30713! ' Elementary Diagram: HV-6217 5 ;
. . .
.
.
}
. -30714- . Elementary Diagram: CCW NCL Supply HV-6212 - 11
- 30716 7 Elementary Diagram
- Surge Tank Isol HV-6225 ' 5- .
- 30717,. Elementary Diagram: Surge Tank isol HV-6505 . ]
- .30718 Elementary Diagram: NSW to Surge Tank HV-. '9 l 6273/6278
'
. 30719~ - Elementary Diagram: NCL lsol HV-6218
-
.
'
i 30733 - Elementary Diagram: CCW Critical Loop A to P025 9
'
HV-6227 i i
. - 30734' Eiementary Diagram: CCW Critical Loop B to P025 8 I HV-6229 ,
!
'
^ 30735~- Elementary Diagram: CCW Miniflow Crossover HV- '5
-
l 6220 30736- :
Elementary Diagram: CCW Miniflow Crossover HV- 7 )
6221 l
- 30737- Elementary Diagram: CCW to Containment HV-6211 13- !
307981 Elementary Diagram: CCW Critical Loop to Letdown 6 HX HV-6293A/6293B 30804 Elementary Diagram: NCL isol HV-6219 13 3086 . Elementary Diagrarn: Reactor Auxiliaries Area 6 Flooding Alarms
30892 Elementary Diagram: CCW Min Flow Crossover HV- 6
. 6551 30883 Elementary Diagram: CCW Min Flow Crossover HV- 6 6552 i
30886 Elementary Diagram: CCW P025 Min Flow to Loop 3 -1 A&B Control l l
.
-16- .,
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. DRAWINGS l
- NUMBER- DESCRIPTION REVISION 30887 Elementary Diagram: CCW Non-Critical Containment - 11
'
Outlet isolation Valves HV-6236 l 30889 - Elementary Diagram: Reactor Aux - CCW Pump 11 Minimum Recirc Valve Solenoids 2HY-6537, -6538, - l 6539 )
30898 Elementary Diagram: Reactor Auxiliary Area Flooding 11 Indication'
31080 Elementary Diagram: Plant Auxiliaries 480V Motor 8 Enclosure Heaters (MCC 2BY)
--31460 Elementary Diagram: PPMU to CCW Makeup Pump 2 P-1018 40127A P&lD: CCW 17 i 401278 P&lD: CCW 27 40127C-- P&lD: CCW 31 40127F P&lD: CCW 27
- 41066, Sh.1-6 Component Cooling Water Pump Tag. No. 2P024 0 (2P025, 2F026, 3P024, 3P025, 3P026) IST Curves 41080, Sh.1-4 Component Cooling Water Makeup Pump Tag. N P1018 (2P1019,3P1018,3P1019) IST Curves LOOP 2FT6243 Loop Diagram: 2FT-6243 (CCW miniflow control) 1 LOOP 2FT6248 Loop Diagram: 2FT-6248 (CCW miniflow control) 1 LOOP 2LT64981 Loop Diagram: CCW Train A Surge Tank Level 0 Sheet 1 LOOP 2LT6498-1 Loop Diagram: CCW Train A Surge Tank Level 2
, Sheet 2
LOOP 2LT7133 Loop Diagram: 2LT-7133 (PPMST level) 1 LOOP'2PT6362-2 Loop Diagram: CCW Pump to Train A HX Pressure 1 !
- [CCW pump autostart)
- LOOP TE6469 Loop Diagram: CCW Temperature 1 P&lD 40150D Condensate Pumps System (Tar.ks) 34-17-
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- MISCELLANEOUS DOCUMENTS
, NUMBE DESCRIPTION REVISION '
' 2101010-98 r --Inservice Pump Test Record - CCW Makeup October 2,1998 i
'
Pump S21203MP1019 -
2P024-09-98 - Inservice Pump Test Record - Component ' September 9,1998 Cooling Water Pump S21203MP024 o I '
l2P025-09-98: ' Inservice Pump Test Record - Component September 15,1998'-
- . Cooling Water Pump S21203MP025
-
DBD-SO23-120 Design Basis Document: 6.9 kV/4kV/480V - 2 System ~ thru DCN 15
'
. DBD-SO23-140 Design Basis Document: Class 1E 125Vdc- .3
,
System :
- DBD-SO23-400 - Design Basis Document: Component Cooling 5 Water -
LER 88-034 Safety Related Component Cooling Water December 15,1988 System Valves Susceptible to Seismically-Induced Common Mode Failures Letter Proposed TS Change to Technical September 4,1998 Specification 3.7.6," Condensate Storage Tank"
- Memorandum Evaluation of Non-1E Interactions February 7,1995
MO- Undervoltage Relays, Sequencing Timers, January 5,1997- i 96030942000 ~ Device and Circuit Test Train A Loss of ' l Voltage Surveillance l None~ Program Response for Generic Letter 89-13, March 29,1991 Service Water System Problems Affecting 1 Safety-Related Equipment None - Memorandum for file - Subject: Testing ' ' October 10,1996
,
Frequency of Component Cooling Water Heat Exchangers in accordance with Generic Letter 89-13
- Report N Spurious Actuation Evaluation, Component ' February 199 ;
M8642 . Cooling Water Gystem Operability ;
Assessment !
-
'RMO .
' Repetitive Maintenance Order for the 50009550000 ' components of loop 2FT-6243 ]
l-18-I i
i