ML20078P926
| ML20078P926 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 09/26/1994 |
| From: | Bowling M, Surface J, Terminella F VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | |
| Shared Package | |
| ML20078P922 | List: |
| References | |
| 94-03-NAPS-A, 94-3-NAPS-A, NUDOCS 9412210028 | |
| Download: ML20078P926 (161) | |
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. North Anna Power StationJ 1 July 11 L29,1994 o
l CNS Report 94-03-NAPS-A e
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l Memorandum
-- - i To J. A. Stall Innsbrook Technical Center From M. L. Bowling September 23,1994 NORTH ANNA SERVICE WATER SYSTEM OPERATIONAL PERFORMANCE ASSESSMENT A Service Water System Operational Performance Assessment (SWSOPA) was i conducted at the North Anna Power Station during July of 1994. The assessment I was conducted in accordance with the NRC Temporary Instruction for Service Water System Operational Performance Inspections, TI 2515/118. The assessment :
team was composed of Corporate and Surry Power Station personnel as well as consultants with previous experience in conducting this type of assessment. l The team concluded that the North Anna Service Water System will perform its required functions during a design basis accident. In addition, it was concluded that the requirements of Generic Letter 89-13 have been implemented. Strengths were identified in the areas of the Service Water restoration project and in corrective action. Several areas for improving Service Water System long term reliability and serviceability to maintain the progress made during the restoration effort were also identified.
The report is attached as CNS Report 94-03-NAPS-A. A summary section is provided at the beginning of the report.
The support of the North Anna Power Station and Nuclear Engineering Services with this evaluation was also greatly appreciated.
M. L. Bowling, Manager Nuclear Licensing and Programs Attachment
t i
I cc: J. P. O'Hanlon :
R. F. Saunders l L. M. Girvin E. W. Harrell !
D. L. Benson E. S. Grecheck L. N. Hartz 5 C. M. Robinson E. R. Smith, Jr. !
J. P. Smith 3 J. A. Stall M. R. Kansler D. A. Heacock W. R. Matthews !
D. A. Christian '
B. L. Shriver J. R. Hayes ;
T. B. Sowers '
P.A.Kemp J. R. Peyton B.K. Day l A. D. Gardner (MSRC)
Team Members: l J. M. Surface -
F. T. Terminella T. A. Kendzia J. E. Lewis D. Wootten .
J. R. Roth J. I. Kelly (SPS) '
K. C. Pier (SPS)
N. S. Turner (SPS) ;
G. L. Prescott (SPS) ,
1 C. R. Bailey (SPS) l i
VIRGINIA POWER SERVICE WATER SYSTEM OPERATIONAL PERFORMANCE ASSESSMENT ,
l JULY 11 JULY 29,1994 NORTH ANNA POWER STATION ,
CNS Report No. 94 03 NAPS-A CORPORATE NUCLEAR SAFETY
/ te h3 Submitted By: /J. M. Surface 7' Team Leader ,
ZZ M F. T. Terminella Assistant Team Leader l Approved by: f ,
M. L. Bowling / Date Manager - NL&P
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA
SUMMARY
AND CONCLUSIONS A Service Water System Operational Performance Assessment (SWSOPA) was conducted at the North Anna Power Station from July 11,1994 through July 29, 1994. The objectives of the assessment were to verify:
- design, operation, maintenance & testing of the Service Water system would assure that its design and licensing bases functions would be accomplished.
- requirements of NRC Generic Letter 89-13, " Service Water System Problems Affecting Safety-Related Equipment" were effectively implemented.
The assessment was performed under the leadership of Corporate Nuclear Safety (CNS) with personnel from Surry Engineering, Surry Operations, Surry Quality Assurance and technical consultants to assure a broad base of design, operation, testing, maintenance and assessment experience. A list of team members is provided in Attachment 1. Team r6sumds are provided in Attachment 6.
The assessment was conducted in accordance with the NRC Temporary Instruction for Service Water System Operational Performance Inspections, TI 2515/118, Revision 1. Tbc ;uethodology is described in Section I of this report.
The details of the assessment scope and conclusions are provided in Section II of the report. The open items and recommendations are provided in Attachment 2. The assessment plan is provided in Attachment 4.
CONCLUSIONS The team concluded that the North Anna Service Water System will perform its required functions during a design basis accident. In addition, it was concluded that Generic Letter 89-13 is being effectively implemented. The extensive restoration work performed to repair degraded piping is considered a strength, but improvements are needed in maintenance and inspection of some Service Water components to ensure long term system reliability, serviceability, and to maintain the progress made during the restoration effort.
The conclusions from each of the five functional areas of the assessment are summarized in the following sections.
l l
l DESIGN:
The overall design of the Service Water System (SWS) was found fully adequate to perform its design functions. The design margins and redundancy provide additional assurance that design requirements will be maintained. However, some Page 1
i
. CORPORATE NUCLEAR SAFETY ASSESSMENT l l I NORTH ANNA POWER STATION SWSOPA E improvements in the single failure reviews and in the major system flow model were identified by the team and are summarized below:
- 1. The results of the SWS single failure review, conducted in accordance with Action IV of Generic Letter 89-13, were not well documented. Some examples of single failures without documented evaluations were found during the assessment, although none of these examples were considered safety significant.
It was also found that the areas of operator error, instrument failure and electrical component failure were not considered during the review.
- 2. Detailed design calculations document the capability of the Service Water System to perform is design function. However, assumptions and inputs to the ,
design calculations were not always current or the most conservative.
- Some inputs to the calculation used to determine spray pond efficiency were not the most conservative.
- Service water heat load calculations were not revised to incorporate a 4.5%
core upgrade and 2% overpower potential.
- The main flow model did not always quantitatively consider the worst-case system conditions when evaluating SW flow to system heat exchangers, although it appears sufficient flow will be supplied.
- Some accident scenarios apparently require isolation of Service Water flow to Recirculating Spray heat exchangers to supply sufficient flow to Component Cooling or Containment Air Cooler heat exchangers. No proceduralized directions are provided to operators to implement this apparent design requirement.
- During normal operations the CCHX throttling could be set up using the
' strong' SW pumps with a ' weak' pump out of service. One of the ' strong' pumps could fail during a CDA. The ' weak' pump would then automatically start. The effect on flow had been considered qualitatively and not quantitatively. The exact flow to the RSHX and other components had not been determined, although it appears sufIicient flow will be supplied.
- The design change used to install shielding around the RSHX radiation monitors in the Quench Spray pump house apparently used a radiation source term to determine shielding thickness that is lower than the Environmental Qualification (EQ) zone radiation fields for a LOCA. A higher background radiatior. field affects the ability of these monitors to detect a leaking RSHX.
i Page 2 l
CORPORATE NUCLEAR SAFETY ASSESSMENT I NORTH ANNA POWER STATION SWSOPA OPERATIONS:
The operation and configuration control of the Service Water System ~was determined to be fully acceptable. Decisions concerning the operability of Service
. Water System components were conservative from a nuclear safety perspective and actions are taken to minimize the out-of-service time for key components. No significant Service Water problems were attributed to operator performance.
A weakness was identified that involved procedural directions given for plant personnel to perform tasks in the Quench Spray house basement during a DBA.
There was no caution that radiation fields in this area could suddenly increase by several orders of magnitude.
1 MAINTENANCE:
Maintenance was found adequate to maintain good material condition in most cases and the implementation of the Service Water Preservation Project is considered a i
strength. Piping condition is monitored and then is replaced or repaired when projections are made that minimum wall thickness will not be maintained until the next outage. Preventive Maintenance is performed on most equipment at appropriate intervals. However, concerns were identified in the areas of material condition, documentation and inspections. These are summarized below:
- 1. Inspections of the service water pumps were not planned or scheduled on a 10 year interval as recommended by the Reliability Centered Maintenance (RCM) Program. The first major overhauls of the pumps were performed in 1983 after approximately 6 years of operation and no other overhauls have been done since that time.
- 2. The SW Pump traveling screens were in a degraded condition. A Work Order was written in 1989 to replace the screens but the replacement had not been planned.
- 3. A program had not been developed to inspect and monitor the condition of the protective coatings on piping and pump casings or to inspect the RSHXs.
- 4. The documentation of as-found and as-left conditions of components inspected for GL 8913 lacked detail.
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CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA SURVEILLANCE & TESTING:
The surveillance and testing program was determined to meet current code requirements and demonstrate the acetptable performance of individual components. However, the following weaknesses were identified:
- 1. Check valves on Service Water Chemical Addition System piping that were intended to isolate this non-safety related line were not being tested. The valves also failed testing done after the issue was identified by the team. Leakage from this line had been qualitatively determined not to have any significant impact on SWS flow. ,
1
- 2. Motor operated valves (MOVs)in the SWS may not be tested under the worst case differential pressure conditions, such as the initial inrush of water into an empty Recirculation Spray Heat Exchanger in dry lay-up. Since the MOVs under question are capable of operating at the greater torques, there should be no safety concern with this issue and this will be examined further.
QUALITY ASSURANCE AND CORRECTIVE ACTION:
The overall Quality Assurance and Corrective Action program was considered effective as indicated by the general acceptable performance of the SW system and the relative scarcity of repeat problems. A programmatic feature that identifies and flags repeat problems is considered a strength as well as the overall engineering support for the SWS. ,
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CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA TABLE OF CONTENTS CONTENTS PAGE
SUMMARY
AND CONCLUSIONS 1
- 1. INTRODUCTION AND SCOPE 5 II. REPORT DETAILS
- 01. Mechanical Design and Configuration Control 7 A. Significant Issues B. Assessment topic discussion J
- 02. Operations 34 A. Significant Issues B. Assessment topic discussion
- 03. Maintenance 45 A. Significant Issues B. Assessment topic discussion
- 04. Surveillance and Testing 63 A. Significant Issues B. Assessment topic discussion
- 05. Quality Assurance and Corrective Action 74 A. Significant Issues f B. Assessment topic discussion III. GENERIC LETTER 89-13
SUMMARY
78 l ATTACHMENTS: i
- 1) Assessment Team 86 ,
- 2) Detailed Recommendations for Open Items 87 ;
- 3) Question Sheet Summary 92
- 4) Assessment Plan 97 l
- 5) Reviewed Document List 104 i
- 6) Assessment Team R6 sum 6s 117 l l
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1 CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA l
SUMMARY
AND CONCLUSIONS l
A Service Water System Operational Performance Assessment (SWSOPA) was l conducted at the North Anna Power Station from July 11,1994 through July 29, i 1994. The objectives of the assessment were to verify: 1 l
- design, operation, maintenance & testing of the Service Water system would !
assure that its design and licensing bases functions would be accomplished.
+ requirements of NRC Generic Letter 89-13, " Service Water System Problems Affecting Safety-Related Equipment" were effectively implemented.
The assessment was performed under the leadership of Corporate Nuclear Safety (CNS) with personnel from Surry Engineering, Surry Operations, Surry Quality Assurance and engineering consultants to assure a broad base of design, operation, testing, maintenance and assessment experience. A list of team members is provided in Attachment 1. Team r6 sum 6s are in Attachment 6.
The assessment was conducted in accordance with the NRC Temporary Instruction for Service Water System Operational Performance Inspections, TI 2515/118, Revision 1. The methodology is described in Section I of this report.
The details of the assessment scope and conclusions are provided in Section II of the report. The open items and recommendations are provided in Attaclunent 2. The assessment plan is provided in Attachment 4.
CONCLUSIONS The team concluded that the North Anna Service Water System will perform its required functions during a design basis accident. In addition, it was concluded that Generic Letter 89-13 is being effectively implemented. The extensive restoration work performed to repair degraded piping is considered a strength, but improvements are needed in maintenance and inspection of some Service Water components to ensure long term system reliability, serviceability, and to maintain the progress made during the restoration effort.
The conclusions from each of the five functional areas of the assessment are summarized in the following sections.
DESIGN:
The overall design of the Service Water System (SWS) was found fully adequate to perform its design functions. The design margins and redundancy provide additional assurance that design requirements will be maintained. However, some Page 1 y -- - - - - - - - - - - - - -
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CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA improvements in the single failure reviews and in the major system flow model were identified by the team and are summarized below:
- 1. The results of the SWS single failure review, conducted in accordance with Action IV of Generic Letter 89-13, were not well documented. Some examples of single failures without documented evaluations were found during the assessment, although none of these examples were considered safety significant.
It was also found that the areas of operator error, instrument failure and electrical component failure were not considered during the review.
- 2. Detailed design calculations document the capability of the Service Water System to perform is design function. However, assumptions and inputs to the ,
design calculations were not always current or the most conservative.
Some inputs to the calculation used to determine spray pond efficiency were not the most conservative.
Service water heat load calculations were not revised to incorporate a 4.5%
core upgrade and 2% overpower potential.
The main flow model did not always quantitatively consider the worst-case system conditions when evaluating SW flow to system heat exchangers, although it appears sufficient flow will be supplied.
- Some accident scenarios apparently require isolation of Service Water flow to Recirculating Spray heat exchangers to supply sufficient flow to Component Cooling or Containment Air Cooler heat exchangers. No proceduralized directions are provided to operators to implement this apparent design requirement.
During normal operations the CCHX throttling could be set up using the
' strong' SW pumps with a ' weak' pump out of service. One of the ' strong' pumps could fail during a CDA. The ' weak' pump would then automatically start. The effect on flow had been considered qualitatively and not quantitatively. The exact flow to the RSHX and other components had not been determined, although it appears sufficient flow will be supplied.
The design change used to install shielding around the RSHX radiation monitors in the Quench Spray pump house apparently used a radiation source term to determine shielding thickness that is lower than the Environmental Qualification (EQ) zone radiation fields for a LOCA. A higher background radiation field affects the ability of these monitors to detect a leaking RSHX.
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CORPORATE NUCLEAR SAFETY ASSESSMENT I NORTH ANNA POWER STATION SWSOPA OPERATIONS:
The operation and configuration control of the Service Water System was determined to be fully acceptable. Decisions concerning the operability of Service Water System components were conservative from a nuclear safety perspective and actions are taken to minimize the out-of-service time for key components. No significant Service Water problems were attributed to operator performance.
A weakness was identified that involved procedural directions given for plant personnel to perform tasks in the Quench Spray house basement during a DBA. i There was no caution that radiation fields in this area could suddenly increas+ by !
several orders of magnitude. I MAINTENANCE:
l Maintenance was found adequate to maintain good material condition in most cases and the implementation of the Service Water Preservation Project is considered a strength. Piping condition is monitored and then is replaced or repaired when projections are made that minimmn wall thickness will not be maintained until the next outage. Preventive Maintenance is performed on most equipment at l appropriate intervals. However, concerns were identified in the areas of material j condition, documentation and inspections. These are summarized below:
- 1. Inspections of the service water pumps were not planned or scheduled on a 10 year interval as recommended by the Reliability Centered Maintenance (RCM) Program. The first major overhauls of the pumps were performed in 1983 after approximately 6 years of operation and no other overhauls have been done since that time.
- 2. The SW Pump traveling screens were in a degraded condition. A Work Order was written in 1989 to replace the screens but the replacement had not been planned.
- 3. An program had not been developed to inspect and monitor the condition of the protective coatings on piping and pump casings or to inspect the RSHXs.
- 4. The documentation of as-found and as-left conditions of components inspected for GL 89-13 lacked detail.
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CORPORATE NUCLEAR SAFETY ASSESSMENT I NORTH ANNA POWER STATION SWSOPA SURVEILLANCE & TESTING:
The surveillance and testing program was determined to meet current code requirements and demonstrate the acceptable performance of individual components. However, the following weaknesses were identified-
- 1. Check valves on Service Water chemical addition system piping that were !
intended to isolate this non-safety related line were not being tested. The valves i also failed testing done after the issue was identified by the team. Leakage from I this line had been qualitatively determined not to have any significant impact on SWS flow.
- 2. Motor operated valves (MOVs)in the SWS may not be tested under the worst case differential pressure conditions, such as the initial inrush of water into an empty Recirculation Spray Heat Exchanger in dry lay up. Since the MOVs under question are capable of operating at the greater torques, there is no safety concern with this issue and the station agreed to examine the issue further.
QUALITY ASSURANCE AND CORRECTIVE ACTION:
The overall Quality Assurance and Corrective Action program was considered effective as indicated by the general acceptable performance of the SW system and the relative scarcity of repeat problems. A programmatic feature that identifies and flags repeat problems is considered a strength as well as the engineering support for the SWS.
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'l . CORPORATE NUCLEAR SAFETY ASSESSMENT l !
l NORTH ANNA POWER STATION SWHOPA l ;
I. INTRODUCTION AND SCOPE ;
A Service Water System Operational Performance Assessment (SWSOPA) was conducted at North Anna Power Station from July 11 to July 29,1994 by i Corporate Nuclear Safety. Team members had expertise in mechanical engineering design, operations, maintenance, testing, and corrective action. A ;
list of team members is provided as Attachment 1. ~
t The entrance meeting was held at NAPS on July 11,1994, to overview the plan i and to establish the leads for the Station interface. An exit briefing was held !
with members of the Station Management on July 28,1994. ;
Section II of the report provides the details of the assessment, All suggested corrective actions for specific issues discussed in Section II are in bold type.
These recommendations are summarized in Attachment 2.
A. ASSESSMENT SCOPE AND OBJECTIVE The objective of the SWSOPA was to conduct assessment of the North i Anna Service Water System to verify that the system design, operation i and performance meets design basis and regulatory requirements.
The scope of the SWSOPAis defined in the Assessment Plan, Attachment ;
- 4. The SWSOPA conducted at North Anna met the requirements of the
, NRC's Temporary Instruction for Service Water System Operational Performance Inspections, TI 2515/118, Revision 1. This required ;
verification of: l
- 1. Thermal and hydraulic performance in the operation of all Service ,
Water System equipment and configuration control in accordance with i the engineering design bases (TI-01.02). i
- 2. Operational controls, maintenance, surveillances, testing, Quality :
Assurance, corrective actions, and personnel training to assure the Service Water Systems are capable of performing its safety-related i function (TI-01.03). '
- 3. Actions taken in response to Generic Letter 89-13 (TI-01.01). :
1 B. ASSESSMENT METHODOLOGY ;
The assessment was conducted using both programmatic (horizontal) and vertical slice methods. The vertical slice was typically performed for review of Design Change Packages (DCPs) and Engineering Work l Requests (EWRs). The overall methodology was to review functional !
areas, investigate and document findings and then look for common i threads that reflect on the programmatic aspects of the review. The '
vertical slice approach allowed for the determination of the completeness of the related reviews and upgrading of engineering reference material, ,
such as revising the UFSAR, drawing control, and the engineering review l
and safety analysis process. {
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CORPORATE NUCLEAR SAFETY ASSESSMENT l NORTH ANNA POWER STATION SWSOPA E The overall approach to the SWSOPA consisted of a review of the Updated Final Safety Analysis Report (UFSAR), Technical Specifications, DBDs, flow diagrams, design drawings, design calculations, equipment specifications, and vendor manuals to establish the design basis requirements for the service water system and its support systems. These items were subsequently compared against the operating, maintenance, and surveillance test procedures at the site.
The SWSOPA evaluated the service water system and interfaces to the safety-related and non-safety-related systems that support the service water system.
I C. ASSESSMENT CONDUCT In preparation for the SWSOPA, team training sessions on reference i material in the Assessment Plan (Attachment 4) were conducted.
Additional training was provided by the North Anna Power Station (NAPS) Training Staff using the same material provided for the Licensed ,
Operator Requalification (LORP) Program. j l
During the assessment, Question sheets were written by the team to i request information or state a position on a potential issue after an initial review of a subject. Each assessment team member was assigned a 1 Station contact to coordinate obtaining responses to the Question sheets.
In total,173 questions were written during the field phase of the assessment. A synopsis of the Question Sheets is included in Attachment 3.
During the course of the assessment'a standard terminology was used to express the level of significance of a review area. A subject was considered under " review" ifinformation was being gathered and assessed and no l conclusions had been generated. A subject was considered in the
" question phase if the review determined there was some significance to a topic under review and more information needed to be provided. A subject could also be determined to be an " issue"if the review indicated there was a technical or compliance issue of some significance.
A Question sheet was considered "open" by the Team if the Station's response did not adequately address the concern raised on the question sheet and further action was considered necessary. Items were designated ;
as " accepted" if corrective actions or discussion in the responses were '
considered adequate. Corrective actions for " accepted" items are being tracked by the Station and are subject to CNS verification until 1 completion. )
1 Suggested corrective actions for each open item and a reference to applicable discussion in the body of the report are in Attachment 2.
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CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA l !
E ;
The overall approach to the SWSOPA consisted of a review of the Updated Final Safety Analysis Report (UFSAR), Technical Specifications, DBDs, . ;
flow diagrams, design drawings, design calculations, equipment specifications, and - vendor manuals to establish the design basis
- requirements for the service water system and its support systems. These ;
items were subsequently compared against the operating, maintenance, !
and surveillance test procedures at the site. :
The SWSOPA evaluated the service water system and interfaces to the safety-related and non safety-related systems that support the service ,
water system.
C. ASSESSMENT CONDUCT In preparation for the SWSOPA, team training sessions on reference -
material in the Assessment Plan (Attachment 4) were conducted.
Additional training was provided by the North Anna Power Station 1 (NAPS) Training Staff using the same material provided for the Licensed !
Operator Requalification (LORP) Program. '
During the assessment, Question sheets were written by the team to i request information or state a position on a potential issue after an initial i review of a subject. Each assessment team member was assigned a Station contact to coordinate obtaining responses to the Question sheets. !
In total,173 questions were written during the field phase of the !
assessment. A synopsis of the Question Sheets is included in Attachment
- 3. .
During the course of the assessment'a standard terminology was used to !
express the level of significance of a review area. A subject was considered under " review"ifinformation was being gathered and assessed and no conclusions had been generated. A subject was considered in the !
" question" phase if the review determined there was some significance to a i topic under review and more information'needed to be provided. A subject 1 could also be determined to be an " issue"if the review indicated there was a technical or compliance issue of some signiScance.
. A Question sheet was considered "open" by the Team if the Station's l
response did not adequately address the concern raised on the question L sheet and further action was considered necessary. Items were designated
! as " accepted" if corrective actions or discussion in the responses were considered adequate. Corrective actions for " accepted" items are being tracked by the Station and are subject to CNS verification until completion.
i Suggested corrective actions for each open item and a reference to applicable discussion in the body of the report are in Attachme,nt 2.
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CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA II. REPORT DETAILS
- 01. MECHANICAL SYSTEMS ENGINEERING DESIGN REVIEW &
CONFIGURATION CONTROL A. SIGNIFICANT ISSUES
- 1. Sinale failure review documentation & comnletenema
- Loss of flow from non-safety related service water piping
- Evaluation of operator error, instrument failure, and electrical component failure
- 2. Calculational Innues
- Conservatism in reservoir spray performance calculations .
- Quantification of flow model worst case analysis I
- Strong pump / Weak pump calculational basis
- Incorporation of changes from core uprate )
1 2
B. ASSESSMENT TOPIC DISCUSSION 01.a. SW dealan nannmntions. boundina conditions. and modmine comnliance with Iicensina commitments & reanlations.
thermal & hydraulic nerformance reanirements consistency of drawina & nrocurement snecifications.
The team reviewed documentation related to the design bases for the service water system (SWS), including calculations, design specifications, engineering reports, design drawings, and portions of the Service Water System Design Basis Document (DBD). Applicable portions of the North Anna UFSAR were also reviewed to identify licensing commitments related to the SWS. The design documentation reviewed confirmed that the design of the SWS was in accordance with the plant licensing basis and regulatory requirements. However, the team did have some comments on the design analyses related to the hydraulic and thermal performance of the SWS, which are summarized below.
Thermal Performance of Service Wate'r Reservoir and Sorays Calculation ME-062, Revision 0, dated 8/12/85 (with addenda through 5/1/87), " Reservoir Performance Analysis," evaluated the performance of the SW reservoir sprays and bypass system during normal operation and DBA conditions. Five computer codes Page 7
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA described in NUREG-0733, " Analysis of Ultimate-Heat-Sink Spray .
Ponds," were employed in the calculation. The team reviewed this '
calculation and identified instances where design inputs used in the analyses were potentially not the most conservative.
For example, SPRCO is the first of five computer codes used in the calculation. This code produces a set of coefficients used in an equation for the thermal efficiency of the sprays. The independent variables in the equation are dry bulb temperature, wet bulb temperature, sprayed hot water temperature, and wind speed. The code calculates an array of spray efficiencies, each corresponding to a set of values of the independent variables. Each set is randomly generated based on ranges of the variables (e.g., DTDRY0, TWETO, and RTW) that are inputs to the code. The code then fits the set of coefficients to this array by a regression method. The team found some input variables and variable ranges used in the analyses produced spray coefficients that would not be applicable during periods of most limiting meteorology. For example:
RTW is the range of wet bulb temperatures considered in developing the equation for the pond spray efficiency. The lower limit of wet bulb temperature used is TWET0. Since accuracy during periods of reduced cooling capability (i.e. higher wet bulb temperatures) is desired, 50*F is sufficiently low for TWET0. The i calculation sets RTW equal to 35'F producing .a maximum i considered wet bulb temperature of 50 + 35, or 85'F. However, i cooling capability varies inversely with wet bulb temperature, and i hot days with 100% relative humidity occur in the summertime.
Therefore, RTW could be input at 45'F to permit consideration of operation of the reservoir during very humid days (e.g., a 95'F wet )
bulb temperature). l l
DTDRY0 is the minimum difference between dry bulb and wet bulb temperatures considered in the analysis. The input value should be zero to include the possibility of days of 100% humidity when spray efficiency will be low. The analysis used 20*F which restricts the analysis to days of favorable meteorology.
4 PHI, ALEN and WID model the passage of the wind through the reservoir's sprayed volume with the wind speed multiplied by sin (PHI). PHI is the heading of the wind with respect to ALEN, the length of one axis of the pond, and WID is the length of the other axis. As air progresses through the sprayed volume,it picks i up moisture, rendering it less efficient in cooling. The calculation I set PHI equal to 90*. However, it did not model this ;
meteorological elTect such that the lowest spray efficiency is '
produced. The calculation modeled the wind normal to the long j axis of the reserenir. Wind direction parallel to the long axis will i yield the lowest spray efficiency. Accordingly, PHI sliould be 90*,
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4 CORPORATE NUCLEAR SAFETY ASSESSMENT .I i I -
NORTH ANNA POWER STATION SWSOPA l )
i
.with WID equal to the length of the long pond axis and ALEN equal to the length of the other axis.
These ranges should ' consider meteorology that will result in !
minimum spray cooling as required by Regulatory Guide 1.27,
" Ultimate Heat Sink for Nuclear Power Plants, Rev. 2,1/76 (Ref. t Question #045.0). {
ANSI N45.2.11 requires that the final design shall be traceable to the source' of design input. However, the team also found that there was !
no documented basis for many of the inputs used in the analysis. !
Sheets 20 and 21 of ME-062 list the input values to computer code !
SPRCO. No reference for most of the values on Sheet 21 is given.
The values are stated as " based on NA meteorology and spray system ;
operation."
l The output of the computer code SPRCO affects spray pond i performance and the results of the analysis in two ways. First, the '
calculated coefficients are used by the code SPSCAN to select the ;
time period with conditions that result in minimum water cooling (as ,
required by Regulatory Guide 1.27) from a meteorological data history tape. Second, the coefficients are used in the SPRPND code ,
to evaluate the spray efficiency during each time step in order to I calculate reservoir temperature. The conservatism of the array of. ,
coefficients has an effect on selecting the most limiting meteorology 1 from the historical data tape. I The team found that no value is input for the variable CMIN, defined in NUREG-0733 as the minimum spray efficiency allowed in the pond temperature analysis by computer code SPRPND. Consequently, the variable will assume the default value,0.20. That is, the minimum spray efficiency the code SPRPND will allow is 20%. Regulatory Guide 1.27 requires analysis of reservoir temperature to minimize water cooling. The spray efficiency is calculated by the computer code depending on meteorological parameters. The variable CMIN may have impact on calculated spray efficiencies (which may be less than 20%)(Ref. Question #045.1).
The team also identified questions related to the heat loads imposed on the reservoir sprays in the reservoir performance analysis.
The computer code SPRPND calculates the temperatures in the reservoir under the combined influence of meteorology and i external plant heat loads. The report prepared by Ford, Bacon & l Davis Utah (FBDU), " Service Water Reservoir and Spray System l Performance Testing and Evaluation," dated 2/79, is the source of I the heat loads used in Calculation ME-062. These heat loads were transmitted in a 12/77 letter from Virginia Power to FBDU. .
Mechanical Engineering was unable to retrieve the Ftter or the l
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l CORPORATE NUCLEAR SAFETY ASSESSMENT l '
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. l calculations that developed these heat loads during the assessment.
1 DCP 86-02-1 (10/8/86) documented the upgrading of the core I power rating from 2775 MWt to 2803 MWt (+12 MW reactor l coolant pump (RCP) power for 2905 MWt ), an increase of approximately 4.5%. The DCP states that the then current (Jan.
1985 for Appendix 8-3) containment analysis was performed for )
an NSSS power level of 2910 MW t
. It is further stated that the containment analysis was re-performed for an NSSS power level l of 2910 MWt + 2% to account for uncertainty in core thermal l power (in accordance with Regulatory Guide 1.49). For the re- l analysis, Westinghouse confirms that their long term energy 1 release takes into account a 2 percent increase in decay heat. The l analyses on which the ME-062 heat loads were based were apparently performed by Stone and Webster Engineering !
Corporation (SWEC) at some time prior to the 12/77 letter !
mentioned above. The DCP does not indicate when the re- 1 analysis took place. ;
Since Calculation ME-062 was completed in August,1985, it was '
expected that the heat loads used in that analysis reflect:
The 4.5% core upgrade power increase, and The containment load from a LOCA unit having operated at 102%
of rated power, a shutdown unit having operated at 102% of rated i power, and all the appropriate auxiliary loads from both units !
consistent with Regulatory Guide 1.49 requirements. i 1
In response to the team's question, Mechanical Engineering stated l that "the two percent core uprating increase in heat load is judged to '
have a negligible effect on this maximum value." However, this ;
response refers to the " core uprate" as 2% (rather than 4.5%). An i evaluation of the total power increase of approximately 7% would be l more conservative (Ref. Questions #061.0 and #150.0). l The team issued a question on the SW reservoir related to protection from postulated tornado missiles. Section 9.2.1.3 of the NAPS UFSAR states that all service water system equipment, piping, and valves are tornado missile-protected, with the exception of the spray piping. In addition, Section 9.2.3 of NUREG-0053 (SER), dated 6/4/76, states that "the reservoir spray piping is not designed to withstand tornado missiles." The team requested the documented analysis that demonstrates that the service water system has the capability to safely shutdown both reactor units without the benefit of the spray system (damaged due to tornado missile) considering the Page 10 i
lI CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA worst single active failure, e.g., failure to open a spray system bypass valve.
In response to this question, Mechanical Engineering stated that a total SW system flow rate of only 1000 gpm would be required to bring both units to a " Hot Shutdown" condition. The licensing basis for North Anna specifies that the safe ' shutdown condition is " Hot Shutdown." Consequently, the station could be brought to a safe shutdown condition following tornado missile damage using either the spray system bypass, the Auxiliary SW system or the partially damaged portions of the eight SW spray supply arrays (if available).
This response is acceptable (Ref. Question # 033.0).
In summary the team recommends that the conservatism in the Reservoir Performance Analysis be confirmed considering the items discussed above (Ref. Question #154.0).
Heat Erchanaer Performance Annivais The team reviewed design documentation, e.g., design analyses, engineering reports, and specification data sheets, related to the thermal performance of heat exchangers. Virginia Power Company .
has developed analytical models used to predict the performance of '
the component cooling heat exchangers. Calculation ME-0420, Revision 1, dated 6/1/94, " Component Cooling Heat Exchanger Retubing/ Replacement Study," was performed to document the capability of the NAPS Component Cooling Heat Exchangers (CCHXs) over a range of SW inlet temperatures and a range of tubes plugged. For the DBA case (Sheet 30), a maximum service water temperature of 95'F is used. Mechanical Engineering informed the team that all components supplied service water are evaluated for flow requirements at a service water temperature of 100*F and that an addendum to Calculation ME-0420 is in process to add evaluations of the CCHXs at a service water temperature of 100*F.
No specific issues were identified related to the CCHX model and performance analyses. However, the team found that maximum calculated reservoir temperatures may exceed 100*F when the considerations related to ME-062 described above are addressed.
Consequently, the effect on CCHX performance, as well as other l components (e.g., charging pumps and chillers), and minimum design . I flow requirements, may require further evaluation.
In response to this issue, Mechanical Engineering has stated that the addendum to Calculation ME-0420 will not be issued until these concerns are resolved. (Ref. Question #132.0)
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0 J l CORPORATE NUCLEAR SAFETY ASSESSMENT
'I NORTH ANNA POWER STATION SWSOPA E .
A number of other design calculations related to the SWS design I basis were reviewed. However, except for those discussed in Section 01.e below, no issues were identified for these calculations.
l Ernansion Joint Onalification The SW lines to the RSHXs contain rubber expansion joints to '
accommodate movement of SW piping in the Quench Spray.(QS) pump house basement. The qualification of the expansion joints and other equipment in the QS pump house basement to withstand potential steam from a high energy line break of the steam supply to the Auxiliary Feedwater pump turbine and radiation from Safety i Injection piping was questioned. The response indicated that most of ,
the equipment in the area had been previously evaluated as i acceptable. Engineering indicated that the maximum temperature .
rating of the SW expansion joints was 250*F. Assuming a room temperature of 250 F and SW temperature of 95 F, the expansion joint temperature would remain below the rated value. The qualification of the expansion joints to withstand the potential radiation environment was not discussed in the response. It is recommended that the expanshn joints be evaluated for .
radiation effects (Ref. Question #06v.0). j Service Water Radiation Monitor Senaitivity Following a LOCA, there will be an increase in the radiation levels in . ,
the basement of the Quench Spray (QS) pump house due initially to 1 shine from the containment. Later in the event, the Low Head Safety !
Injection (LHSI) pump suction will swap to the containment sump i and discharge to the RCS and to the suction of the High Head Safety '
Injection (HHSI) pumps. The LHSI discharge linas which provide {
suction to the HHSI pumps pass through the QS pumphouse .
basement. Recirculated reactor coolant containing radioactive - I material from failed fuel in these lines will increase the radiation l levels in the QS pump house basement where the SW radiation monitors for the RSHXs are located.
The doce rate in the QS pump house basement (below grade) could l reach on the order of 10 to the 5th R/hr in the event of a LOCA !
according to Stone & Webster calculation based on the TID-14844 i LOCA source terms, the source terms used for the environmental Qualification (EQ) zone). Shielding has been installed around the RSHX radiation monitors per DCP 80-S33 to reduce background radiation for the monitors. It was not clear in the design change package that the assumed background radiation fully accounted for, the increased radiation levels because the source term used was signiScantly less than the EQ zone source term.
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1 CORPORATE NUCLEAR SAFETY ASSESSMENT l NORTH ANNA POWER STATION SWSOPA l l The team questioned whether the SW radiation monitors would be '
able to detect a faulted RSHX with the increased background :
radiation levels. According to UFSAR Table 11.4-1, the background I radiation level for the RSHX SW monitors should be below 5 mR/hr, l although the response to the question (and the DCP) indicated the minimum sensitivity of 28 mr/hr to Co-60 gammas is all that is required. A rough calculational estimate on the background radiation in the radiation monitor cubicles by the team based on the !
EQ zone source term above did not confirm the background levels ,
were low enough to achieve the minimum sensitivity required. It is recommended that the apparent discrepancy between the EQ zone source terms and the source term used for the shielding calculation for the RSHX radiation monitors be resolved (Ref. l Question #071.0). )
01.b. Constatency of drawinam andr dealan. NRC reanirements.
licenalng commitments.
The team found that in some instances design drawings did not reflect as-built configurations. However, these were deviations with minor significance and are discussed in Section 02.a. !
01.c Dimerennncima between anaratinn and damlan dacmnanta.
The team reviewed plant operations to determine consistency with ,
design documents. During normal operation, four SW-pumps and 2 headers are required to be operable in accordance with T.S. 3.7.4.1.
Otherwise, the SW pump discharge pressure must be above the minimum pressure determined by the SWS flow balance test (currently 58 psig). This is accomplished by throttling SW via the .
Component Cooling heat exchanger (CCHX) outlet valves. This ensures that in the event of a DBA, where only two pumps will -
remain operational given single failure considerations, that sufficient '
SW flow will remain available for the Recirculation Spray Heat '
Exchangers (RSHXs). !
Particular attention was given to the SW Flow Balance conducted each refueling outage. The SW Flow Balance is performed under 1- '
PT-75.6, " Service Water System Flow Balance," and 2-PT-75.6, '
" Service Water System Flow Balance". The reason for this flow balance testing is to verify that the SW flow to the RSHXs meets l
design requirements during a Containment Depressurization '
Actuation (CDA) given that the SWS is common and shared between both units. -
The initial setup for the SW flow balance consists oflining up the SW supply to equipment normally in service during a CDA, such as ,
control room chillers, Instrument Air compressors, CCHXs, and -
charging pump coolers. This is to ensure that the lest closely Page 13 i
n.- ,-,- -, , - -, -n , _ - . _ - . , - . . - , , , . - - . - ~ _ - , - , _ - . _ - - , . . - - _ - . _ _ . _ , . . -- -, . , - . - . _ - ,,, -- , . . , ,
b
-l CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA l l l
l approximates ~ the predicted CDA alignment.' The ' two' worst- .
performing SW pumps, one per header, and all spray arrays are placed in service. The SW system is throttled to the appropriate discharge pressure condition as determined by the previous i performance of the flow balance test. The affected unit RSHX supply and return valves are then opened and flow is secured to the affected units' CCHXs. Flow of at least 4500 gpm to each RSHX is verified.
Further, if flow to each RSHX is greater than 5000 gpm,'then no additional SW system balancing is required. If flow balancing is required, the SW pumps are throttled to 35 psig or greater to allow room for the throttling operation. The RSHX inlet MOVs (SW-MOV-203s) are alternately throttled to obtain at least 5000 gpm per RSHX. l The new valve positions are recorded and the throttle positions are marked. The flow balance test is then completed by isolating the RSHXs and establishing flow to the affected unit CCHXs. The discharge pressure of the SW pumps is recorded to determine the new flow balance criteria for future operation. The SW system is i then returned to a normal alignment. 1 1
The sequence and methodology of the SW flow balance test allows for !
postulated events occurring during the test. For example, when the l RSHX valves are unisolated, the J bus RSHX supply valves (SW-MOV-101B&D) and the H bus return valves (SW-MOV-105A&C) are opened in order to allow remote isolation in the event ofloss of one ,
emergency bus. l The following inconsistencies were identiSed:
Flow To HXs on Non-accident Unit - The System Design Basis Document (SDBD) for the SW system specifies that during a Loss-of-Cooling-Accident (LOCA) with Loss-of-Offsite Power (LOOP) scenario, the SWS is capable of providing 7000 gpm to the CC . -
heat exchangers on the unaffected unit from time 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A review of the most recently completed flow balance periodic tests <
identified that the as-left SW flow to the CC heat exchangers was l less than 7000 gpm.
In response, Mechanical Engineering stated that the 7000 gpm )
was based on a conservative SW inlet temperature based upon j four nuclear units in operation and a DBA on one unit. At '
approximately one hour into the accident only two RSHXs would be required with the flow from the other two RSHXs capable of being diverted to meet the non-accident unit CC requirements.
Mechanical Engineering concluded that there is adequate SW flow available for all component cooling water design requirements and the SDBD will be revised to correctly state the required SW flows.
However, Emergency Operating Procedures (EOPs) do not contain guidance for isolating SW to the RSHXs when Recirculation Spray l
Page 14 1
1 i
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._,__,-,,,..__~_,.-,_mm,__ _.,. .m_.+_,. ,,-, ___.- . _ _ _ _ _ _ _ , . _ .m +,~,,_.,mm_-.,.-
44- Yf+
, .y l
- CORPORATE NUCLEAR SAFETY ASSESSMtNT l '
.l' NORTH ANNA POWER STATION SWSOPA I i
is secured. The team concluded that, if credit is taken for isolating SW to the RSHXs during the course of the DBA, procedures should be revised to include directions 'for these actions (Ref. Question #081.0).
Flow To Containment Air Recirenlation Coolina on Non-accident' :
.Umt - The SDBD for the SW system specifies that during a LOCA l with LOOP scenario, the SW system will provide 1200 gpm to Containment Air Recirculation Cooling of the unaffected unit from 30 minutes to at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following event initiation. These '
coils are normally cooled by the Chilled Water System. The SW flow balance tests, PT-75.6, do not account for this flow.
Therefore, during the event, the flow balance to the RSHXs could >
be affected by this change in system flow distribution. There is no
~
documented analysis that determines the flows supplied to safety 1 related components when these coolers are aligned to the SWS.
In response, System Engineering stated that the SW flow balance assures design basis flows to the RSHXs at the initiation of the ,
event. It was further stated sufficient SW will be available 30 minutes into the event by isolating RSHX(s). Mechanical ,
Engineering indicated that Emergency Response Team personnel would determine the effects on SW system performance resulting from manually aligning the SW system to supply flow to the recirculation air coolers during a CDA. As discussed above, EOPs i do not implement an apparent design requirement for isolating ;
SW to the RSHXs when Recirculation Spray is secured (Ref.
l Question #151.0). As discussed in below and in Section 01.e, CCHX throttling ultimately assures flow is sufficient. .
SW system throttiing is accomplished by operating procedure 0-OP- [
49.6, " Service Water System Throttling Alignment." This procedure '
determines whether SWS throttling is required and provides the necessary guidance and instruction to accomplish the task. A review of operating procedures for placing other significant SW loads into service was also performed to ensure that the SWS is appropriately i throttled after the load is placed into service. The following procedures were identified:
AP-35, Loss of Containment Air Recirculation Cooling," provides instructions for responding to a loss of Containment Recirculation Cooling caused bs loss of chiller units, fans or Chilled Water flow.
In the event that no Chilled water is available, such as during a LOOP, AP-35 directs the operator to align SW to the Containment Recirculating Air Fans in accordance with OP-21.1, " Containment Ventilation." OP-21.1 directs the operator to perform 0-OP-49.6 if SW adjustments have been made.
Page 15
l CORPORATE NUCLEAR SAFETY ASSESSMENT I NORTH ANNA POWER STATION SWSOPA
- 0-AP-27, " Malfunction Of Spent Fuel Pit System," provides instructions for responding to various Spent Fuel Pit (SFP) malfunctions. SWis used es a backup to normal SFP cooling. AP-( 27 does not require performance of 0-OP-49.C following placing or removing SW from service. 0-AP-27 should be revised to require performance of 0 OP-49.6 when SW is placed in service or removed from service for the SFP (Ref. Question
, #091.0).
t Operating logs were also reviewed. With the exception of an error found in 1-LOG-4A, "CRO Surveillance Sheets (Modes 5 & 6)," all
( CRO operating logs were found to contain the correct information i related to SW throttling. The log listed a minimum SW pump l discharge pressure of 53 psig rather than 58 psig. (Ref. Question
- 031.0) DR N-94-0984 was submitted to correct the identified discrepancy.
01.d. Sinale active failure vninarabilities: Imnact on interfacina avstema. notential common mode failures.
In response to Action IV of GL 89-13, Technical Report ME-0026 was written to document the Single Failure Review for the North Anna Service Water System. No major issues were identified during the review of ME-0026, however, a weakness was found in the general nature of the report which did not document the specific components that were reviewed and did not document the findings of the review.
Also there was no evidence that a single failure review was done of components that are cooled by SW in the back up mode.
The report states that electrical redundancy, diversity, and separation were not "re reviewed." During the course of the assessment, it was identified that failure of a single relay (IX-ISWN08) in the SW pump logic could result in the simultaneous l l start of two pumps on one emergency bus during LOOP sequencing if l an Auxiliary SW pump is in service. This will affect the transient response of the EDG during loading and ultimately overload the EDG at the end of the sequence. Mechanical Engineering responded that this scenario is bounded by a minimum safeguards condition which considers loss of an emergency bus. This is true, however, the team ,
was unable to locate a procedure where this function of the relay is '
tested and its failure would then not normally be detected. This one relay represents a challenge to an entire train of safety systems (Ref.
Question #035.0)
Another example was examined that describes a scenario where the l expected system flows could be changed while running the Auxiliary Service Water (ASW) pumps. A single failure of active components 1-SW-MOV-119 or 2-SW-MOV-219 (ASW pump discharge cross connects) with backflow through tvfo check valves (these check valves Page 16
-l CORPORATE NUCLEAR SAFETY ASSESSMENT l NORTH ANNA POWER STATION SWSOPA are not tested and therefore could be undetected failures) can divert some SW flow from emergency equipment to the circulating water !
intake structure. Mechanical Engineering states that this is not a r credible scenario in that only one single failure has to be taken and the single failure would be used to eliminate the SW water pumps necessitating the use of the ASW pumps. This, however, is not
, necessarily so because the ASW pumps could be running during r testing or for reservoir make-up when the event begins, although the l length of time this occurs is a small fraction of the total running time of the SWS (Ref. Question #100.0). This was later found not to be a concern because these flowpaths are procedurally isolated during *
, testing and makeup (discussed in more detail in Section 01.e, Topic:
" Flow Balancing").
Other potential single failures documented in Question #100.0 were !
identiSed. Loss of flow from this piping was not accounted for in the flow model.
l Inlet and outlet piping for the safety related CCHX radiation j monitor 1-SW-RM-107 and its pumps was designated as non- ;
safety related.
Inlet and outlet piping for the SW discharge radiation monitor 1- l SW-RM-108 and its pumps was designated as non-safety related.
Normally open, non-safety related Polymer Addition System piping was isolated by a non-safety related check valve.
It should be noted that corrective action was initiated for all of these !
items.
Although none of the single failures would compromise the SWS design basis function, the need to better document what was '
reviewed and evaluate any potential impact was demonstrated.
Mechanical Engineering agreed that Technical Report ME-0026 will t be revised to more fully document the SWS components evaluated, :
and to specifically document the evaluation of single failure due to operator error, instrument failure, or electrical component failure. In addition, it is recommended that components in the SW flowpath when it is the designed back up cooling source be included in the single failure review (Ref. Question #100.0).
Page 17
CORPORATE NUCLEAR SAFETY ASSESSMENT :
NORTH ANNA POWER STATION SWSOPA 01.e. Blofonlina and siltina. Features to detect flow loss. Flow hainneina. Pumn run-out. Minimum and marimum limits for valve nositions.
Flow Balance Calculation ME-0309, Revision 0, dated 4/30/92, Base Case KYPIPE2 Service Water Flow Model for North Anna Power Station was i performed to model the NAPS service water system. Calculation ME-0327, Revision 0, dated 12/15/92 (and Addenda A & B), Service Water ,
Operational Base Case KYPIPE2 Flow Model uses operational data ,
as indicators to re-set the analytical base case flow model. The team '
reviewed both of these calculations and found the calculations generally do not always address worst case conditions for all modes of operation of the service water system considering all single failures.
Specific weaknesses are identified as follows:
Isolation of SW chemical addition - In the event of a seismic disturbance with a loss of offsite power, non-seismic lines are subject to pipe breaks which may result in diversion of flow from safety-related components to the break. DC-85-48-3 added the service water chemical addition system. Section 8 of the Safety Analysis for this design change describes the consequences of a pipe break in the non-seismic portion of this piping which connects to the service water headers. The safety evaluation states that the loss of SW inventory due to a break in this piping has been included in the reservoir performance analysis, ME-062.
The team reviewed Calculation ME-062 and found that the indicated 144,000 gallon loss is based on an estimated flow of 100 gpm. However, there is no documented basis for the 100 gpm break flow used in Calculation ME-062 (See Section 01.h). The team independently estimated that flow from a break of this size pipe would significantly exceed 100 gpm used in the analysis assuming a header pressure of 58 psig. Consequently, flow to ,
some components may be slightly affected by the flow diverted to the break.
The response to this question indicated the Chemical Treatment Feed line break flow of 100 gpm represents a small (0.33%)
percent of total SW flow supplied and the effects of this loss on reservoir heat rejection capability were addressed in Calculation ME-062. Although this is true, flows to some safety-related components themselves represent a small percentage of total flow and may be affected. For example, the SW DBD (Table 6.1-4) indicates the control room chillers require a flow of 237 gpm.
However, the flow modelincludes a fixed demand flow of 200 gpm (which does not satisfy the 237 gpm minimum flow re,quirement).
It has not been not been quantitatively determined what the Page 18
CORPORATE NUCLEAR SAFETY ASSESSMENT l NORTH ANNA POWER STATION SWSOPA impact would be on this flow (and flows to other components) by ,
the loss of flow through the break as well as flow losses (or diversions) discussed below, although this has been qualitatively evaluated and a determination has been made by Engineering that there is sufficient flow to all heat loads. 1 This issue is discussed further in Section 04.e. In response to :
Question #002.0, the station indicated the potential loss through these lines would be controlled by testing the installed check valves and putting the manual isolation valves in the IST program until an analysis can be performed that incorporates the losses. When this analysis is performed, the magnitude of the flow loss should be re-examined and the effect on the SWS be evaluated for potential impact (Ref. Questions #072.0, #073.0)
- Auriliary Service Water Subsystem - During operation of the Auxiliary Service Water Pumps (lake-to-lake mode), check valves '
1-SW-23 and 1-CW-10 prevent reverse flow which could divert flow back to the circulating water intake structure through the 8-inch line and the screen wash pump. However, these check valves are not in the IST program and are not tested in reverse flow.
Consequently, no credit can be taken for the valves performing this function since they could be in an undetected failed condition (e.g., stuck open disk or no disk at all). The response to inquiries in this area states the Auxiliary Service Water pumps are strictly a backup subsystem and are not taken credit for in a design basis accident.
The team agrees that the Auxiliary Service Water pumps are a backup source of cooling water as required by Regulatory Guide 1.27 and no credit is taken for them in a DBA. However, in NUREG-0053 (SER), the NRC concluded that the North Anna
" ultimate heat sink design conforms with Regulatory Guide 1.27, and is acceptable." The SER further stated that:
"The design conforms with Regulatory Guide 1.27 with regard to the availability of two sources of water and redundant aqueducts,"
and "The spray system is subject to damage as a result of tornado missiles that may reduce its cooling efficiency. However, the Lake Anna source of cooling is not subject to such damage."
Further,in response to questions related to the effects of tornado missiles on the spray piping, Mechanical Engineering stated that, although unlikely, the " station could be brought to a safe shutdown condition following tornado missile damage" using the Auxiliary Service Water pumps. In addition, the SW DBD (Section 3.3.5) states that the Auxiliary Service Water pumps could be used following a safe shutdown fire. The team concluded that during these scenarios or when the ASW pumps are in Page 19
CORPORATE NUCLEAR SAFETY ASSESSMENT l l NORTH ANNA POWER STATION SWSOPA E service for testing or makeup, flow could possibly be diverted from the SWS back to the circulating water intake. But because required SWS flows for a shutdown are significantly lower that for an accident condition, at a minimum a check for gross reverse flow through this line should suffice. It was found that two governing operating procedures, 0-OP-49.3, " Service Water Reservoir i Makeup," and 0 OP-49.2, " Service Water System Lake-to-Lake ,
Operation," and the ASW periodic pump tests,1/2-PT-75.5, contain steps to isolate the screenwash pumps when operating the.
ASW pumps, precluding any leakage. Therefore, no concern exists i for this topic (Ref. Questions #072.0, #033.0).
Valve 2-SW Valve 2-SW-14 is a normally open valve in a 1" -
line off the supply header downstream of pump 2-SW-P-1B to a corrosion monitoring station. A restricting orifice is located in the non-Q piping downstream of the valve. However, in a seismic event, no credit can be taken for the integrity of the downstream ,
piping. The calculation does not account for the loss of flow through this line. Flow from this valve would be in addition to that from the pipe break in the event of a seismic disturbance ;
since it is a continuous flow (Ref. Question #072.0). '
Moderate Energy Cracka - Assuming operation with only three SW pumps and the CCHXs throttled in accordance with Technical Specification 3.7.4.1, failure of a diesel generator during a loss of offsite power could result in loss of a SW header. This would leave two pumps operating on the remaining header. The calculation ,
did not evaluate the effect of a moderate energy crack in the :
remaining header which could divert flow from safety-related '
components. The response to inquiries on. this topic (Ref.
Question #072.0) indicated the SW header is designed to seismic, safety related criteria and is not considered to experience this type of failure (" moderate energy crack"). In this case, the
" moderate energy crack" is the initiating event and is not ;
considered a " failure," according to NUREG-0800, Section 3.6.1-and it associated Branch Technical Position on high and moderate ,
energy line breaks. However, an examination of the SW DBD '
indicates there is no commitment was made to the Standard !
Review Plan (NUREG-0800) for the North Anna SWS.
- Containment Recirculation Cooling Sunnly - The SWS also provides backup for the containment recirculation air cooling coils and the spent fuel pool coolers. The calculation has no provisions ,
to model flow to these components. Mechanical Engineering informed the team that Calculation ME-0322, Revision 0, dated. l 11/17/93, " Service Water MOV Operating Torque Requirements,"
provides the capability to model these components. However, tids calculation was developed to determine worst case maximum differential pressures across MOVs in the GL 89-10 program.
Page 20 l
CORPORATE NUCLEAR SAFETY ASSESSMENT .lE I- -
NORTH ANNA POWER STATIO,N SWSOPA Since the intent of this calculation was to develop these maximum >
MOV differential pressures, it is not clear that the scenarios evaluated addressed worst case conditions that would minimize ,
flows to safety-related components or that results obtained were evaluated to determine if adequate flow is supplied to all safety- !
related components. The calculations do not account for potential !
variations of flow to heat loads during these conditions, and there i is no quantitative analysis to address these issues. :
The response to inquiries in this area indicated SW supply is only - !
a back up capability for the recirculation air cooling coils and the spent fuel pool coolers. Flowing SW to recirculation air coils .
would be a decision made by the Emergency Response Team personnel, and they would have to determine the effects on the SW system at that time (Ref. Question #072.0). The team found that procedures permit aligning the recirculation air coolers at :
any time and do not provide direction to isolate the coolers on a !
CDA. This has a potential consequence on the non-accident unit, i since the coolers on the accident unit will automatically isolate on a CDA. Consequently, no credit can be taken for isolating the coolers on the non-accident unit on a CDA. However, as i mentioned in Section 01.e, the CCHXs are throttled by procedure !
to ensure the required header pressure for flow to the RSHXs if f the coolers are valved into the SWS. The fact that the spent fuel cooler procedure for the same condition does not contain similar instructions had already been mentioned in Section 01.e. Pudher, i if four pumps are operable (3 running after a single failure) then ,
this is not an issue even without the throttling because there is l more than enough flow supplied for the extra load.
+ Pnmn Curves - No basis or reference source is provided in the calculations for the input used to establish the service water or auxiliary service water pump curves. Mechanical Engineering informed the team that the input was derived from data obtained from pump testing. The input represents a " composite curve" for all SW pumps tested based on the data obtained. However, the method used to establish this input is not documented. Further, the input pump curve does not account for maximum degradation i in pump performance permitted by the In service Testing '
Program for these pumps (i.e., IST limits). ANSI N45.2.11 requires that the final design shall be traceable to the source of design input. The basis for the SW pump curve used and the methodology used to develop the curve should be documented (Ref. Question #072.0).
- Fixed Flows - The calculations use fixed demand flows totaling 600 gpm to the charging pump lube oil coolers, instrument air compressor coolers, and control room chillers. Consequently, it is not clear that flows to these components may not be higher or Page 21
CORPORATE NUCLEAR SAFETY ASSESSMENT
{ NORTH ANNA POWER STATION SWSOPA lower than the fixed demands used. Mechanical Engineering informed that team that Calculation ME-0295, Revision 1, dated 12/8/93, Verification of Adequacy of Piping Diameter to Control ,
Room chillers, confirmed that the control room chillers are !
supplied adequate flow. However, this calculation did not determine a flow distribution in the SWS. The calculation determined that more than sufficient head is available to aatisfy i design requirements. Consequently, this branch may affect flow :
to other safety-related components. Similarly, it is not clear that !
flows to the other branches (modeled as fixed demands) may not i actually affect flow from other components. In addition, some sort .
of upper limit to acceptable leakage back through the SW pumps' discharge check valves has not been established.
Engineering considers application of the fixed demands in the model is appropriate given the known margins in the systems based on operational experience and operating data and that small differences in flow will not impact the CCHXs minimum expected flow. However, as also stated above under the discussion of the SW chemical addition system, the team found that these small differences have not been quantitatively evaluated (Ref.
Question #072.0). i l
The team concluded that the cumulative effects of these items may !
effect SW flows to safety-related equipment. In response to these 4 concerns, Mechanical Engineering stated that the hydraulic model as !
documented does provide for the worst case scenarios and that the calculations were not intended to address every line of supply. I However, the issues identified in this topic should be i evaluated for potential impact on SWS flows (Ref. Question
- 072.0). A sensitivity study would be one way to evaluate the effect '
of changing and fixed flows. t "Strone" Piimn/" Weak" Piimo i
The team identified an additional failure scenario that had not been ;
evaluated for impact on flow distribution during a CDA:
During flow balance testing the CCHX outlet valve is throttled to supply design flows to the RSHXs. This testing is performed with the worst (" weakest") two SW pumps selected (See Figure 1, Operating Points 1 and 3) to assure adequate flow is supplied to the RSHXs in the event of an a CDA assw dng the failure of a SW pump with one pump taken out of service. .t is conceivable that the two " strongest" pumps could be operating normally with one (weak) pump taken out i of service. This would require re-throttling the CCHXs to a pump discharge pressure of 58 psig with these strong pumps aligned and operating (Figure 1, Operating Point 2). However, since this ,
adjustment would change the total system resistance, flows to safety- !
Page 22 i
l CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA lg related components would be redistributed differently from that established during flow balance testing.
Subsequent to re throttling the CCHXs, with the system operating normally, one of the strong pumps could fail on a CDA. The CDA would auto-start the remaining weak pump (Figure 1, Operating Point 4), potentially affecting flow to the RSHXs and other safety .
related components for this pump alignment with the altered CCHX throttle valve positions.
In response to these questions, the station provided an evaluation of the potential for runout during this scenario. The response also 4 stated that design flow of 4500 gpm to the RSHXs is still achieved for this scenario. However, no documented analysis to support this statement (i.e., that design flow to the RSHXs is achieved) was provided in the response. Further, based on IST permissible limits for SW pump performance, the worst pump performance could be as much as 7% degraded, and the best pump performance could be as much as 2% better than an established reference. Consequently, it is conceivable that the difference between pumps could be as much as 9%. The team found that there was no documented analysis that '
quanti 6es flow to the RSHXs and other safety related equipment for this case, but there is no evidence to contradict the station's engineeringjudgment that sufficient flow is available.
l Mechanical Engineering has stated that the issues raised in this scenario will be evaluated and documented (Ref. Question #064.1).
Biofouling The SWS has a sub system, the Service Water Chemical Addition -
System, designed to inject several different chemicals to mitigate the effects of biofouling and corrosion. Details of the chemical control program are addressed in Section 03.d. It should be noted that the SW Chemical Injection System is designed for the relatively low feedrates inherent in a recirculating system. The equipment cannot ,
(and was never intended to) deliver the feedrates necessary to treat the once through mode oflake-to lake. Due to the lack of use of the lake to-lake mode, this does not appear to be needed. Chemical residuals can be re-established after the SWS is returned to its !
normal mode of operation.
Pnmo Runout For the additional failure case described in the " Strong" Pump /" Weak" Pump topic above, the team found that there was no documented analysis to demonstrate that sufficient service water ;
pump NPSH and submergence exists to assure pump runout limits ,
are not exceeded considering-l Page 23
CORPORATE NUCLEAR SAFETY ASSESSMENT
- NORTH ANNA POWER STATION SWSOPA maximum calculated SW pump flow based on the potential difference between pump performance curves for this scenario and highest reservoir water level.
maximum calculated reservoir temperature during a postulated DBA.
i These factors should also be evaluated as recommended in the flow balancing docussion (Ref. Questions #064.1, 072.0, and #132.0)
SW Pomo Start Throttlina Section 4.9 of Procedure 0-OP-49.6, Revision 3-P1, SW System Throttling Alignment, states that to " minimize pressure surges in the system, the SW Pump Discharge Valve should be throttled when SW Pumps are started."
The team requested the documented basis for this instruction, the magnitude of these " pressure surges," and questioned how throttling would be effected during an SI or CDA. In addition, the team requested the documented analysis to evaluate potential water hammer effects during system transients, e.g., loss of offsite power with subsequent pump restart.
]
In response to these questions, Engineering stated:
- Surge magnitude was not measured.
The basis for the instruction was " good operating practice."
The intent of the instruction was to preclude leakage across the RSHX inlet and outlet MOVs attributed to pump startup pressure surges. j
- Leakage past the valves has been minimal sinco corrective action was taken two to three years ago.
- No transient analysis exists for the SWS. However, no water hammer events have been observed during CDA testing with two SW pumps started unthrottled and two SW pumps running.
Engineering further stated that the relatively slow opening times of the RSHX valves and the relatively large volume of air in the RSHXs act as a SW system dampener to surges.
Based on this response, it appears that the original cause of problems l that required pump throttling has been corrected. The station could I possibly stop the throttling since it may no longer be necessary, but l
Page 24 l
i
)
CORPORATE NUCLEAR SAFETY ASSESSMENT l I- NORTH ANNA POWER STATION SWSOPA I throttling is a normal operational practice and operation in either manneris acceptable.
Water Hammer Annivsis-As mentioned above, the team found that there is no analysis of the ,
effect of potential water hammer. Water hammer has occurred in !
SWSs at other sites despite the presence of slow opening MOVs. The transients imposed during pump trip and subsequent startup, e.g.,
during a loss of offsite power, may allow sufficient time for draining high points in the system that precipitate column separation. This ;
may occur at locations other than the RSHXs (e.g., the chillers, l charging pumps, or instrument air compressor piping). Upon i subsequent pump restart, separated columns rejoin and may produce pressures in excess of design pressure. Further, ti e interaction of l the various branches cannot easily be assested without a '
comprehensive analysis. Testing usually does not fully duplicate these conditions, nor is it desirable to perturb the system in this ,
manner during a test. !
Examination of the NAPS SWS indicates its configuration precludes this scenario. The highest elevation in the system is at the SW !
reservoir, which essentially acts as a " head tank." Static pressure alone from the reservoir provides sufficient pressure to prevent column separation. Further, since the return lines go back to the reservoir, there is no credible drain path that could also allow column separation. Additional analysis does not appear to be indicated in '
this case (Ref. Question #046.0).
01.f. Elooding from SW leaks.
The team reviewed documentation of features that mitigate flooding from SW system leaks or leaks from other systems.
The review confirmed that the design provides adequate protection from SW leaks or leaks from other systems. Documentation reviewed included design specifications, engineering reports, operating procedures, abnormal operating procedures, and the flooding sections of the IPE and IPEEE.
UFSAR Appendix 30 contains an analysis of the consequences of postulated pipe breaks outside the containment. The postulated break of a high energy line was shown not to negate the functions of any structures or systems important to safety, and not to negate any redundancy of any system or component required to operate as a result of the postulated failure. Although flooding protection is considered adequate overall, some concerns were identified as follows: ,
Page 25 l
l CORPORATE NUCLEAR SAFETY ASSESSMENT l l NORTH ANNA POWER STATION SWSOPA l Imnlementing Flooding Evaluation Results In 1989, Technicon Enterprises Inc. (TEI) performed an evaluation of i the impact of NRC IN 87-49, " Design Deficiencies in Outside Containment Flooding Protection". It also included a review of the concerns raised by INPO SOER 85-5. This evaluation was focused on moderate-energy pipe (with a through-wall leakage crack as opposed to complete pipe rupture). The evaluation considered the potential impact of flooding by water sources resulting from failures of equipment outside the area being investigated as well as from the inadvertent release of water in the area created by possible personnel errors during maintenance operations.
TEI concluded that all areas are adequately protected from potential flooding effects created by moderate energy pipe breaks, leakage from external sources or personnel errors. To ensure that installed water collection and pumping systems continue to remain effective, TEI made several recommendations. Of these, all were verified as implemented except the following:
- Sump pumps be cycled periodically
- Floor drains be flushed periodically In the Virginia Power cover letter of this report, Engineering concurred with the TEI findings and recommendations.
In response, Mechanical Engineering concluded that this recommendation was not required for the Emergency Switchgear &
Instrument Rack Rooms because automatic sump pumps were installed as recommended by TEIin this report. This was considered an enhancement to an already adequate system. Mechanical Engineering did not address the above TEI recommendations for the installed water collection and pumping systems in all other areas reviewed. A performance problem in this area was discovered during assessment team walkdowns where the turbine building SW valve pit was found to contain water due to previously undiscovered failures of three components (See section 03.a for more discussion). The team concluded that the TEI flooding recommendations should be implemented (Ref. Question #097.0).
Service Water Valve House The Service Water Valve House (SWVH) was not identified as having equipment important to safety even though there are safety . clated MCCs located in the SWVH. Consequently the SWVH was not analyzed for flooding in this report.
Page 26 i
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA Mechanical Engineering supported this position since the only potential for floodmg in the SWVH is from SW piping and components which are safety related and seismically designed. The TEI evaluation focused on the potential impact of flooding due to moderate energy pipe through-wall cracks and flooding by personnel errors during maintenance evolutions. The team concluded that the SWVH should be evaluated to the same criteria used by TEI to ensure that flooding does not adversely impact any safety related equipment in the area (Ref. Question #102.0).
01.g. Salamle an=11ficatinn of unfatv rainted sections and isolation of non safetv-related sectionp.
The team identified several cases where non-seismically designed piping was not adequately isolated or the isolation feature was not adequately tested. Consequently, a break in the non-seismic piping could result in the diversion of service water flow or loss of reservoir inventory, or both. The effects of these cases on service water flow balance is discussed in detail in Section 01.e above.
During the team's walkdown of the SWS, the SW pump house was toured. Unit heaters were observed in various locations of the pump house. For example, a heater was suspended over a safety-related panel near a SW pump. The team questioned if these heaters are seismically supported. A falling heater could damage nearby safety-related equipment.
In response to the team's question, Engineering stated that the Unit Heaters, 01-HV-UH-15A through 15E and 36A&B are non-safety-related, non-seismic components. While not seismically designed, the anchorage for these unit heaters is expected to survive a design basis seismic event. Engineering further stated that the seismic ruggedness of these components will be evaluated under the NAPS -
SQUG program to officially document this preliminary evaluation (Ref. Question #006.0).
During the walkdown of the Unit 2 chiller room, a drain line from the overhead room was examined. The team questioned if the line were seismically designed, since it runs above all three chillers and was supported by hangers not typically associated with seismic designs.
As with the unit heaters mentioned previously, the lines are expected to survive a seismic event and will be formally evaluated under the SQUG program (Ref. Question #051.0).
Page 27
l CORPORATE NUCLEAR SAFETY ASSESSMENT l NORTH ANNA POWER STATION SWROPA 01.h. SW modifications: 50.59. compatibilltv with dealan basma. i mvision of procedures.
The team reviewed portions of a number of Design Changes and Engineering Work Requests and identified several questions and concerns. For example:
SW Chemient Addition Svatem Modincation DC-85-48-3 added the SW chemical addition system. Section 8 of the Safety Analysis for this design change describes the consequences of a pipe break in the non-seismic portion of this piping which connects to the SW headers. The safety evaluation states that the loss of SW inventory due to a break in this piping has been included in the reservoir performance analysis, ME-062. The technical merits of this analysis are discussed in detail under " Flow Balancing" in Section !
01.e. The Safety Analysis took credit for daily monitoring of the supply line pressure by station personnel in order to identify and terminate the leakage within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period from onset. At present, no procedure or log requires such monitoring. Therefore, the '
assumption that the break flow will be terminated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from onset cannot be credited. Procedures should be revised in ,
order to perform the daily monitoring of the chemical !
addition system non seismic piping added by DC 85 48 3 as required by the safety analysis (Ref. Question #073.0). ;.
In addition, the Controlled Document / Review and Revision (CDRR) for DC 85-48-3 indicated that revision to the North Anna Setpoint Document (NASD) was required to add various level switches, pressure switches and relief valves. The NASD does not currently ;
contain these components, however.
Engineering indicated that the final Tech Review for the DCP was never completed because there was no mechanism in place to drive the DCP for close-out. VPAP-0103, " Design Change Process," did not exist at the time DCP 85-48-3 was completed. A final Tech Review will be performed to ensure DCP close-out is completed and applicable programs are revised as required (Ref. Question #147.0).
Control Room Chiller Service Water Strainer ModiHcation DCP 92-123, " Replacement of Main Control Room Chiller Service Water Strainers," replaced the self cleaning strainers associated with -
the SW supply to the Main Control Room Chillers with stainless steel wye-type strainers due to increasing corrective maintenance. The North Anna Setpoint Document was not revised to delete the differential pressure switches associated with the Unit 2 strainers.
In addition, NCRODP-36, " Secondary Plant Ventilation Systems,"
was found to still contain a description of the automatic straineis.
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l 1
l CORPORATE NUCLEAR SAFETY ASSESSMENT lg
{- NORTH ANNA POWER STATION SWSOPA ,
l i
Engineering was unable to determine why the NASD was not revised even though the Station Setpoint Coordinator had issued Interim '
Notification of Setpoint Changes No. 93 9-5. The Service Water operator training manual was also not revised during preparation ;
and review of the DCP. A new CDRR has been generated to correct this (Ref. Question #146.0). !
All other associated documentation related to both DCPs mentioned !
above were adequate. The discrepancies identified relating to . t program / document revisions indicate that previous review / revision ;
processes for DCPs were not entirely effective. The current design change process has a review process Intended to prevent such discrepancies.
l Other Par kwes l
In addition to the above, the following packages were reviewed:
DCP 85-46-3, " Service Water Corrosion Rate Monitoring" [
JCO 92-05, " Evaluation of Service Water Piping Integrity For Concrete Encased Piping" !
The 50.59 evaluations for these packages were reviewed and found to ,
be adequate. For these changes, requirements for maintenance, revision to operating procedures, training and periodic testing have :
been identified and implemented where necessary.
DCP-86-02-1 documents upgrading the core power rating from 2775 :
MWtto 2893 MW t(+12 MW RCP power for 2905 MWt ). The team '
. reviewed portions of this DCP in conjunction with the review of Calculation ME-062, Reservoir Performance Analysis (See Section 01.a above, Thermal Performance).
01.1. Action IV of GL 89-13. '
In response to Generic Letter 89-13, Action IV, North Anna Power i Station has completed the following actions:
The Service Water System Design Basis Document, SDBD-NAPS-SW, Revision 01, dated 12/31/93 was issued. The team reviewed portions of this document and found it to be comprehensive and detailed. The Open Items for the DBD are listed in Chapter 24.0. <
These Open Items identify questions, missing documentation, and -
inconsistencies that require resolution. The team re, quested the Page 29
I I I
[ . CORPORATE NUCLEAR SAFETY ASSESSMENT l i
l NORTH ANNA POWER STATION OWSOPA I documented evaluations for these Open Items that identify any associated safety concerns and their resolution, or the rationale supporting the position that no safety concerns exist for them.
In response to this request, " Minutes of SDBD Open Item Review Meeting for SDBD-NAPS-SW, Service Water System, North Anna Power Station," was supplied. Page 1 of this document indicates the minutes review and discuss the open items identified for Revision 1, :
of the SW DBD. However, the team found that this document does .
not include evaluations for all Open Items identified in Revision 1 of the SW DBD. Two of the Open Items not evaluated were questioned due to potential impact on the design of the SWS. For example:
Open Item 24.2.11 identifies several issues, including:
Containment pressurization analysis for a LOCA., which supports the current technical specification for containment temperature :
and pressure, is based on a constant SW temperature of 97'F, not on the " design maximum SW temperature" of 110*F. Design documents demonstrating satisfactory operation of the containment recirculation air coolers with 110*F SW temperature have not been identified.
Open Item 24.6.3 identifies similar issues involving an assumed constant SW temperature for accident analyses rather than the time dependent SW temperature, and SW temperatures less than calculated used in these analyses.
The team requested the documented evaluations that demonstrate that there is no safety concerns for these and all other Open Items identiSed in Revision 1 of the SW DBD. Engineering indicated that there was an error in printing the Summary Report and the list of Open Items was for Revision 0. It was further stated that during the revision process, none of the Open items were deemed to have any significant safety implications. The Open Item Management System database will be updated within 90 days of the " issuance date," July 12,1994 (the date the revision to the SW DBD is issued by Records Management) rather than 90 days from the date noted on the DBD, December 31, 1993. Although the timing of the formal documentation of the new Open Item Evaluations comd be shortened, the new Open Items were examined during the revision procest of the DBD and any item with obvious safety implications should have been identi6ed and acted upon (Ref. Questions #07.0, #07.1).
The technical issues in the two open items mentioned above were assessed by the team (Ref. Questions #046.0, #046.1, #132.0). The team confirmed the items did not have safety significance.
Single failure reviews have been performed. Technical Report ME-0026 is the Single Failure Review for the North Anna Service Water Page 30
l CORPORATE NUCLEAR SAFETY ASSESSMENT I NORTH ANNA POWER STATION SWSOPA System in response to Generic Letter 89-13. However, detailed results of these reviews were not documented. This issue is discussed in detail in Section 01.d.
Walkdowns were performed under the Configuration Management Program. A number of discrepancies were noted during these walkdowns and Drawing Update Requests (DURs) were generated to l correct these items. See Section 02.a for further discussion of the results of the assessment team's walkdowns.
01.j, Monitorina system nerformancer trendina. enaineerina evaluation. onerability determinations, Monitoring system degradation and performance trending is addressed in Section 04.
01.k. Conalatency between ninem actuation setnoints and desian basis / assumptions.
Five alarm setpoints were selected at random for examination. The design basis for 4 out of five could not be found in the Setpoint Document and questions were submitted on those setpoints (Ref.
Questions #106.0,115.0,124.0, & 125.0). It was later discovered '
these items were in identified in Section 24 of the SW DBD and thus had been previously identified as open items and will be resolved as <
part of that process.
i t
a Page 31
l i
@@ ~
134 l
(58 psi) i 1 2S C I i IV/lS I I I b
i
' l l
i' 2V i I i
= OW O
O'l@
FIGURE 1 Pumo Curves:
2W - Combined pump curve for two " weak" pumps operating.
2S - Combined pump curve for two " strong" pumps operating.
1W/1S - Combined pump curve for one weak and one strong pump operating.
System Resistance Curves:
C- Flow Balance Testing simulating CDA with CCHXs throttled to supply required flows to RSHXs A- Post Flow Balance Testing establishing 58 psi pump discharge pressure B- After re-throttling CCHXs to 58 psig pump discharge pressure with two " strong" pumps operating normally ,
Page 32
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA FIGURE 1 (cont.)
D- CDA alignment with re throttled CCHXs Ooeratine Points:
3- Operating point when flow balance is established with CCHXs throttled to supply design flows to RSHXs during flow balance testing with two weak pumps.
1- Operating point with 58 psig pump discharge pressure established with two weak pumps for operation with one pump taken out of service.
2- Operating point with two strong pumps operating during normal operation and CCHXs re-throttled to 58 psig (one weak pump take out of senice).
4- New operating point with one weak pump and one strong pump operating during CDA resulting from failure of one strong pump and auto-start of remaining weak ptunp.
l Page 33
CORPORATE NUCLEAR SAFETY ASSESSMENT l I' gP.TH ANNA POWER S' RATION SWSOPA E
- 02. OPERATIONS A. SIGNIFICANT ISSUES -
1 Opnnideration of Radiation Fields in Quench Snrav House 1.
4 Bgsement durinar DBA B. ASSESSMENT TOPIC DISCUSSION 02.a. Svatam walkdown for comnarison with demian drawines.
r The accessible portions of the SW system outside of the containments were walked down to verify accuracy of design drawings. Thirty-one drawing discrepancies were noted, The types of discrepancies included: vent and drain valves installed and labeled not on drawings, vent and drain valves installed with no labels not on drawings, instrumentation and instrument isolation valves installed t and labeled not on drawings, and slight differences in physical configuration compared to the drawings (Ref. Questions #023.0, 025.0, 027.0, 029.0, 043.0, 082.0, 085.0, & 089.0). Several of these discrepancies had been previously identified by the station and drawing change requests (DCRs) had been submitted previously.
DCRs were submitted by the station for those discrepancies not '
previously identified. There is a backlog of DCRs to be incorporated into station drawings. A priority system has been established to control the order of DCR incorporation into drawings based on signincance.
Labeling, in general, was easy to use with mark number and noun description consistently identiSed. A number of the component labels :
in the field were missing (16), incorrect (9), or confusing (4) (Ref.
Questions #019.0, 024.0, 026.0, 028.0, 030.0, 039.0, 040.0, 044.0, 083.0, 084.0, 088.0, & 0123.0). The label problems were similar to the drawing discrepancies described above. Some of the missing labels had been previously identified and were waiting new labels.
Label problems identified by the team which were not previously identified were put into the corrective action process by the station (incorrect labels were removed and temporary pink tags put on were i
appropriate).
The ability to operate the SW system in accordance with the procedures and system design requirements was evaluated due to the labeling discrepancies and the differences between actual installation and the drawings. The ability to operate the SW system in accordance with the normal and emergency operating procedures was not affected. The only affect of these discrepancies on the operation of the system would be during maintenance, or venting and draining of portions of the SW system. The drawing and labeling Page 34
CORPORATE NUCLEAR SAFETY ASSESSMENT
- l. NORTH ANNA POWER STATION SWSOPA discrepancies would be identified during normal maintenance through the normal work control and tagout programs and should prevent these discrepancies from causing problems. Since this is a common system in continuous service during either unit operation, venting and draining would only occur under unusual maintenance ,
activities which would be subjected to extra controls. The refurbishment of the SW piping is one of these unusual maintenance activities, and the controls used have prevented the drawing discrepancies from causing any problems. The team recommends that outstanding SWS DCRs be processed in a timely manner (Ref. Question #153.0).
Walkdown of the control room revealed that several meter indications (flow, pressure, amps, etc.) had " candy cane stick-on" markings used by operations to indicate normal system parameters. . These .
markings are not considered operator aids by the station and are not .
strictly controlled. It was also noted that the discharge pressure for
- the SW pump markings were not consistent between the pumps and were not consistent with the as left requirements of the operating procedures. The existence of these markings outside the operator aid program, and the specific differences for the SW ;
pump discharge pressure compared to the operating procedures should be evaluated (Ref. Question #099.0).
02.b. SW ops. ARPs. APs conalstency with desian and imnlementability. Adenuncv of flow instrumentation.
temnerature and flow monitoring.
Procedure Review Results i l
General comments about categories of SW procedures are as follows: l Emergency Procedures - All of the emergency procedures which require SWS manipulation were reviewed. The abnormal weather condition procedures, loss of reservoir and lake level procedures, flooding procedures, annunciator procedures related to the SW components, and the operating procedures for the safety-related portions of the SW systems were reviewed. In general, the procedures were of good quality, provided the operators with the necessary direction to operate the equipment, and implemented the design requirements (one exception to this is discussed in Section 01.e).
Emergency Ooerating Procedures - The Emergency Operating l Procedures (EOPs) orovide instructions to check that the SW l pumps start and SW motor operated valves (MOVs) reposition j when required. No direction to check SW system parameters (flow, discharge pressure, pump amps, etc.) is provided. The station stated that the system parameter checks are' required by l 1
l Page 35 i
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA training and practiced in the simulator (IIef. Question #110.0).
Abnormal responses of the SWS should also be identifiable through control room annunciators, although, as mentioned below, the annunciator response procedures are limited in addressing various accident condition failures. The EOPs do not address termination of SW flow to the RSHXs that are no longer required to be in service or other adjustments to SW flow. Design implications of this have been addressed in Section 01.c.
Annunciator Resnonse Procedures - The annunciator response (AR) procedures generally provided direction for addressing the normal cause of the annunciator and depend on the operators to correctly diagnose and respond to off normal initiating events.
The review found these procedures are not necessarily consistent as to the amount of diagnoses performed or detail of directions, and do not address limitations imposed by the problem.
Comments to this effect were made on the individual question sheets and are generally considered enhancements (Ref.
Questions #094.1, 094.2,128.0,133.0,134.0,135.0,136.0,137.0, 138.0,139.0,140.0, 141.0,142.0, 144.0, 148.0, & 149.0).
However, one annunciator response procedure had actions to check local system conditions of the RSHX SW radiation monitoring sample pumps may not be possible due to an intense radiation field during an accident. Although the response to the question indicated this annunciator response procedure would not be used during a DBA, it is actually most likely to be used during '
an accident because this is the only time these radiation monitors are required to function. It is recommended that ARP 1K D4 be revised to include precautions about entering a high radiation zone (Ref. Question #143.0).
The station identified that these procedures were scheduled to be upgraded as part of the Virginia Power upgrade process. Several operators were interviewed as to the adequacy of the annunciator response procedures and they did not consider them to be inadequate. A cursory review of station Deviation Reports did not indicate a performance problem with using the annunciator response procedures.
Periodic Tests - The procedures for SW components that required operations manipulation of components were accurate and provided good direction. These procedures do use the OP ,
procedures for a lot of the manipulation of operations components, ,
such as starting SW pumps. Interviews with the operators. l revealed that they thought that at times there were too many ,
procedures required to be entered to perform one test evolution. A I cursory review of the station deviation reports did not reveal any i
Page 36
.I CORPORATE NUCLEAR SAFETY ASSESSMENT lg 1 l NORTH ANNA POWER STATION SWSOPA j problems caused by having to use multiple procedures for one test l evolution. t
?
- Operatina Procedures - Normal Operating Procedures in the 48 :
and 49 series address various aspects of SWS operation. These ,
procedures specify throttling of the SW pump discharge valves prior to pump start, which is discussed in Section 01.e, Topic, "SW Pump Start Throttling." T1.ese procedures have long precaution and limitation sections which may be difficult to absorb and remember while the procedure is in progress. Standard pump checks prior to starting and after starting of the SW pumps are :
not required by procedure, but are implemented as skill of the '
craft practices. .
The procedures normally check SW system parameters after - .
system configuration, and specific values are used in relation to i pump discharge pressures, but not for system flows which are !
checked as " normal". One procedure, which is not often used, would allow mixing of SW and Auxiliary SW pumps which will !
impact crosstieing headers (This is being evaluated by the !
station). These procedures do require a number of operator i actions to ensure adequate SW flow to RSHXs will occur automatically during a DBA, but these actions were found to be 1 necessary. Comments on these procedures were submitted during i the assessment and are considered enhancements (Ref. Questions !
- 057.1, 101.0, 116.0, 122.0, 129.0, & 130.0). !
Several procedures reference other procedures by incorrect numbers I because the procedures have been upgraded and assigned different i numbers. Personnel are aware of this and there is a cross reference :
in DMIS from the old procedure numbers to the upgraded procedure l numbers. References to other procedures are corrected as the i procedures are upgraded or revised. -
l The station is in the process of assigning mark numbers and taggmg ,
instrumentation valves. Personnel are aware of this, and can use the ;
procedures as is without having to PAR in the valve numbers. The !
station informed the team that as these procedures were upgraded or revised, the new mark numbers will be included. :
i Snecific Procedure Comments j Specific procedure comments are as follows: ;
1/2-OP-21.1, " Containment Ventilation," addresses providing SW to the Containment Air Recirculating Coolers. This procedure provides good direction for putting SW in service to the ;
Containment Air Recirculating Coolers and removing it from !
service. However, use of this SW alignment crossties the SW j 4
6 Page 37 I
i
1 CORPORATE NUCLEAR SAFETY ASSESSMENT l-I NORTH ANNA POWER STATION SWSOPA l l headers without implementing administrative controls. The need for additional controls when SW is aligned to the Containment Air Recirculation Coolers and the SW headers should be evaluated (Ref. Question #113.0).
- 1-AP-22.5, " Loss of Emergency Condensate Storage Tank Level,"
has actions to provide SW to the Auxiliary Feedwater (AFW) pumps as an alternate suction source. The capability to supply.
SW to the AFW pumps as an alternate source provides excellent system flexibility. This procedure includes detailed steps on how '
to implement this capability and restore system alignment when no longer required. A question concerning the isolation point for ;
the SW to the AFW, which in effect crossties the RSHX SW headers is addressed in section 2.d. ,
- 0-AP-41, " Severe Weather Conditions," addresses actions ,
necessary for high wind conditions, including tornadoes and -
hurricanes. The procedure provides good direction for protection of the SW system including the temporary conditions caused by ,
the SW restoration project.
- 0-AP-5.1, " Common Unit Radiation Monitoring System," 1-AP-5,
" Unit 1 Radiation Monitoring System," addresses abnormal or ,'
high readings on the common RMs including SW discharge to the lake, SW discharge to the reservoir, and CCHX SW Discharge RM. Attachment 14 provides good direction on how to identify the leaking CCHX. Attachments 15 & 16 provide direction to :
responding to the SW discharge to the lake RM and the SW l discharge to the reservoir RM respectively. The direction in these ;
attachments as to identifying the source of the activity is minimal.
Attachment 15 did not contain instructions to evaluate the dose to the public but will be revised to reference VPAP 2103 for offsite dose calculations (Ref. Question #117.0). The directions in the attachments for response to a RM malfunction are good.
inleakage to the RSHX SW side through normally closed inlet '
isolation MOVs. This procedure opens drain valves which are between the outside containment wall and the RSHX outlet MOVs that serve as the containment isolation boundaries. The .
procedure provides a caution to close the valves if a CDA occurs. l The procedure also closes and verifies these valves at the end of the testing. A question was submitted relating to whether this form of administrative control met the requirements of Technical Specifications to administrath ely control normally closed manual containment isolation valves when opened (Ref. Question #104.0). I The station response was that this method of control was in j accordance with standard practice and did meet their I interpretation of the Technical Specifications. A cursory review of Page 38
CORPORATE NUCLTA SAFETY ASSESSMENT ]g NORTH ANNA PL ; 3R STATION SWSOPA station Deviation Reports for the last two years did not reveal any problems resulting from controlling these valves in this method.
Although no specific problem was found with this method of administrative control it is not as formal as seen elsewhere.
0 AP-27, " Malfunction of Spent Fuel Pit System," provides direction to establish SW flow to the Spent Fuel Pit Coolers for some situations. The procedure provides direction for draining the CC and establishing SW flow through the coolers. The instructions do not address what to with the Waste Gas Compressors and Spent Fuel Cask storage connections which are located within the boundary to be drained of CC and refilled with SW. The procedure does not address checking SW system parameters after valving in this header, to ensure that the CCHX SW throttling is correct. There are no administrative controls established to terminate this SW flow path in accident conditions, which may be necessary by the system design (this is discussed -
further in the design section). There are no directions for restoring the systems to normal after the need for SW flow to the coolers is eliminated. The station has agreed to evaluate this procedure in light of these observations (Ref. Question #091.0).
0-AP-40, " Abnormal Level in North Anna Reservoir (Lake),"
addresses high and low lake level conditions. The only affect on the SW system is to put the SW system in reservoir to reservoir mode of operation (which is the normal mode).
0-AP-35, " Loss of Containment Air Recirculation Cooling,"
includes direction aligning SW to the Containment Air Recirculating coolers if necessary. This is accomplished using a normal operating procedure.
0-AP-47, " Unit Operation During Opposite Unit Emergency,"
includes general guidance to observe the common systems affected, including SW and systems SW serves.
0-FCA-1," Control Room Fire," address an Appendix R fire in the control room. The procedure ensures SW pumps are running or starts them, but does not check system parameters. The procedure also sends an operator to ensure the spray array isolation MOVs are open and to rack out the associated breakers.
The procedure does not require checking the status of the spray array bypass MOVs and if these valves are open the spray array flow will be less than design. For Appendix R, however, a CDA is not coincident with a fire and Engineering indicated that pump run out will not occur with all bypass and spray valves fully open.
Checking the valves does not appear to be needed.
Page 39
.l CORPORATE NUCLEAR SAFETY ASSESSMENT l l NORTH ANNA POWER STATION SWSOPA E
- 0-FCA 9," Service Water Pump House Fire," address an Appendix R fire in the Service Water Pump House. The procedure i addresses the loss of SW through the use of AP-12, and provides the additional instructions necessary for the fire situation.
- 1&2-FR-Z.2, " Response to High Containment Sump Level," has the operator assess where flooding may be coming, including :
cheeks of RSHX SW flow and temperature for " normal" conditiona.
Normalis not defined.'
- 1-AP-5, " Unit 1 Radiation Monitoring System," addresses :
abnormal or high readings on the unit RMs including Discharge Tunnel and RSHX SW Discharge RMs. Attachment 4 provides direction for the Discharge Tunnel RM response. Steps to isolate all potential radioactive releases and to sample the discharge tunnel are given without direction as to how to accomplish the actions. Directions are provided for evaluating the release for '
emergency plan actions and reportability, and what to do for a suspected RM malfunction.
Attachment 10 addresses the RSHX SW outlet RM response. The procedure directs HP to obtain affected HX SW samples and to check radiation levels in the Quench Spray Basement, if the RSHXs are in service. This area contains the LHSI to HHSI containment sump recirculation lines, which could prohibit this ,
action due to high radiation fields from this piping. Ifpersonnel are present in the area when the sump recirculation phase of SI :
1 initiates, they could receive very high doses. Due to the facts that ,'
- 1) containment sump recirculation may automatically initiate, 2)
HP technicians may not be fully aware of plant status during an accident, and 3) that conditions may be rapidly changing, it is recommended that precautions about the potential for a sudden increase in radiation in the Quench Spray -
basement be included in 1/2 AP 5 (Ref. Question #118.0). The directions in this attachment to evaluate and if necessary implement isolating the affect RSHX SW are good.
4
- 0-AP-39.2, " Auxiliary Building Flooding," addresses flooding in
- the auxiliary building. The most likely sources of water for '
flooding this area are the CC, SW, FP, and Chilled Water systems.
There is good guidance to respond to flooding from these sources. i The procedure also checks for and provides direction for response to flooding originating from other sources such as PG or SG :
Blowdown. It was noted that the procedure did not identify separately the likelihood of water originating from the piping tunnel from the turbine building, either due to turbine building flooding or due to a pipe rupture in the tunnel or associated valve
, pit which includes an out of the way portion of the SW system
~
(Ref. Question #094.2).
Page 40
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA a
f 0-AP-39.1, " Turbine Building Flooding," addresses flooding in the turbine building. Flooding in this area could not be initiated from the SW system, but can be initiated from other sources including the CW system. The procedure provides good guidance for
- terminating the sources of flooding. :
0 AP-12, " Loss of Service Water," provides direction for complete or partial loss of SW flow. The procedure provides minimal guidance for. where and what to look for to diagnose the cause, .
and when to do implement steps such as stop the SW pumps. The
, procedure provides excellent detailed direction on performing ,
unusual system lineups to provide alternate cooling to the charging pumps, aligning for use of the Auxiliary SW pumps, cross connecting SW headers and responding to a SW header rupture. The procedure specifies which parameters to observe for >
SW system performance, but uses " normal" for the desired value (Ref. Question #131.0). This procedure is practiced during simulator training.
02.c. SW onerator trainina technical comnlatanaan. accurncv. and incornoration of modifications. ,
All the operators are trained on SW on a annual basis. This includes classroom instruction, simulator instruction, Task Performance Evaluations, and walkdowns/ Job Performance Measures (JPMs) on the system. Five Task Performance Evaluations, one JPM, two simulator scenarios and the SW System Course Qualification requirements were reviewed, and found to provide appropriate personnel training. General training information associated with the !
SW system was reviewed and one mistake .was found that was 1 promptly corrected by the station, and no other discrepancies were :
noted (Ref. Question #001.0). l Two operations crews were observed in simulator training, which !
included a loss of SW scenario. This was a challenging simulator l scenario, which started with high RCS leakage, a loss of containment instrument air, rod control problems, a loss of instrument air resulting in a reactor trip and then a loss of Service Water. Overall the crews' performance and the training were acceptable.
Recent restoration activities associated with the SW system have impacted the operation of the SW system. Operators have been kept well briefed on the ongoing maintenance, and have been involved with the planning and execution of this work.
Interviews with both the licensed and non licensed operator instructors were conducted. Operator knowledge of the,SW system was found to be fully acceptable.
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CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA __
02.d. Imnlementation of valve allanments to safetv-related comnonents. Comnatlhility of normal system alianment with accident conditions. throttle valve nositionina. heat exchanaer flow control due to chanrina temneratures.
The valve lineups for the safety-related portions of the SW systems configure the system as required by design except as noted below. No discrepancies in the valve lineup procedures versus actual configuration were discovered. Virginia Power does not consider the Flow Diagram drawings an approved method for documenting whether a valve is open or closed, only the valve lineup Operating Procedures are used.
Normal system alignment is fully compatible with accident and abnormal conditions except for where SW is lined up to non-safety related systems, and the SW alternate feed to the Auxiliary Feedwater (AFW) system. The procedures aligning the SW system to non safety related systems do not ensure the isolation of these interfaces for various accidents. This issue is addressed in the design section and in the procedures section 01.e, Topic, " Flow Balance," of this report. The SW alternate feed to the AFW system is from the two SW headers to the RSHXs which are normally dry headers.
These two SW headers are effectively crosstied by the current valve alignment which keeps the 6 inch supply valves to the AFW crosstie line open with isolation to the AFW provided downstream. For certain abnormal accident scenarios, such as a LOCA with partial j loss of one emergency bus and a tube leak on a RSHX, this valve '
lineup would require entering a prohibitively high radiation area to isolate a leaking RSHX. It is recommended that keeping the SW to AFW crosstie valves open be evaluated considering the impact on RSHX isolation (Ref. Questions #095.0 & #070.0).
Throttling of SW system valves for system performance'is addressed by the various operating procedures and supported by analysis and 4 l
test results. The throttling of the SW supply to the CCHXs is l accomplished by using SW pump discharge pressure and not based on i CCHX SW requirements. Section 01.e, Topic, "SW Pump Start Throttling," of this report includes discussion of the adequacy of the basis for throttling requirements. Specific comments on the procedures are included in the procedure section. Automatic functions, such as MOV repositioning, SW pump auto start, TV closure, and fail safe positions of flow control valves, are not violated by the normal system alignment.
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CORPORATE NUCLEAR SAFETY ASSESSMENT
[ NORTH ANNA POWER STATION SWSOPA 02.e. Walk-throuah of ops and svatem drawinas with onerators and engineers. Procedure nerformance and accessibility of eaulnment. availahility anti onerability of snecial eeninment.
operator knowledge.
Walk-throughs were conducted on various normal operating and abnormal operating procedures. Licensed and non-licensed operators were interviewed to verify their system, procedural, and actual plant physical knowledge, in regrrds to the SW system. System 1 walkdowns were conducted with various engineers. The operators -
and engineers were knowledgeable of the system, and consulted each ,
other about problems when appropriate.
~
Some of the SW system components are located in valve pits which are not normally accessible, requiring a missile shield to be removed.
These components do not normally require local observation or manipulation. For some of the SW components either keys and/or oxygen detectors are required to get in the areas. These limitations
, appeared to be acceptable for the procedural requirements to operate the associated equipment.
One item of concern was identified in that the alternate indication for reservoir level is a ruled marker mounted on the side of the SW pump house. Since the air supply that supplies the normallevelindication is non-safety related, this indicator may need to be used for taking TS required surveillance data. All the operators interviewed knew the location of the marker, its purpose and that they would need :
binoculars to read it. The concern is that the marker is not legible l below normal water level due to dirt, and the marker is not in a l program to verify that it is marked accurately. It is recommended '
that the marker be verified as being accurate and a plan be developed to ensure it can be read below normal water level (Ref. Question #017.1). l 1
02.f. Onerator technical knowledsre of system oneration. l surveillances. onerability.
l Several operators, ranging from non licensed to licensed SROs, were l interviewed using standardized questions based on required knowledge (i.e. not in the procedures) to: perform operator rounds, ,
i take SW related logs, implement safety related or emergency l procedure stops for SW manipulations, describe actual SW system response to evolutions, be aware of recent operational changes, and know design requirements and functions related to the SW system.
All of these operators had a thorough knowledge of the system and were familiar with the design, operational, and system response I requirements.
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CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA 02.g. Local oneration of couinment for indicatina enuinment.
environmental conditions durina accident conditions.
l All procedures were examined from the standpoint of ability to perform them under adverse conditions when applicable. Factors considered included lighting, habitability (CO2, Halon, Fire, Smoke, j
Water, Radiation, Steam, Temperature, etc.), communications, and )
j knowledge level of the procedure performer. Procedures required to I be performed under accident conditions were acceptable, with the exception of the addressing RSHX SW in certain scenarios. ;
Abnormal and annunciator procedures for sampling the SW from a l RSHX , the ability to check local system conditions of the RSHX SW radiation monitoring sample pumps, and the ability to isolate SW for 1 a RSHX, all may be prohibited by dose rates in certain scenarios. )
These items are discussed further in the 2.b and 2.d sections of this report. i l
02.h. Onerationni controls to nrevent travelina screen cloaaina.
Function of the traveling screens for the Circulating Water (CW) intake structures is not needed for accident conditions. They are not safety-related. Even in the lake-to lake mode, the flows and velocities of the auxiliary SW pumps are well below that of the CW pumps, so those screens should not impede operation even if they are not running.
The reservoir traveling screens in the SW pump house are not considered by Operations to be required for operability of the SW pumps. System design appears to support that the original design required these screens to be functional for the SW pumps to operate. l For example, the screens have safety-related power supplies. In its current condition, however, the screens do not appear to be l technically required. Presently, the reservoir is treated with biocides i to prevent any life from sustaining in the water. The reservoir is l located so that it is not susceptible to wind blown debris. The l design function of the screens should be reassessed, either downgrading them to non safety, removing *. hem, or requiring I them to be operable (Ref. Question #004.4).
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. CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA i
- 03. MAINTENANCE A. SIGNIFICANT ISSUES ,
i
- 1. Service Water System MaterinI Condition
- SW traveling screen material condition
- Separated piping from polymer addition line to CR chiller SW
- 2. Inanections and Documentation .
- Documentation of as found/as left condition
- Documentation & timeliness of SWS chemistry corrective action
- Planning and scheduling of SW pump inspections -
- Guidance / training for heat exchanger and pipe inspections
- Inspection of RSHXs
- VPAP-0811 compliance B. ASSESSMENT TOPIC DISCUSSION ,
The team reviewed the material condition of Service Water system !
components and the maintenance practices on selected components. Work history was reviewed to identify components requiring frequent corrective maintenance. Selected completed Work Orders (WO) were reviewed for completeness, corrective actions taken, and documentation of as found and as left conditions. Ongoing maintenance was also observed during the ,
assessment. .
' 03.a. Walkdown for materini condition.
During the walkdowns the overall material condition of the SWS and surrounding areas was good. Station cleanliness, lighting, and control of work areas appeared to be excellent. Standards were usually consistently implemented. The SW Chemical Addition Building and equipment contained in the building were found to be in ,
excellent condition, even though the building is in a remote location and the equipmentis non-safety related. One exception to the overall ;
condition was at the reservoir (SW Pump House and the SW Valve House), which is also remote from the controlled area. Although the material condition at the reservoir was acceptable, the number of minor problems discovered indicated standards were not as rigorously applied as inside the controlled area (Ref. Questions u020.0, #021.0).
The walkdown of the Control Room Chiller Room found work requests tags previously submitted by station personnel. These WRs were on bolting that lacked full thread engagement of several expansion joints in SW lines to the control room chillers. During Page 45
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CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA l
E
[
examination of other SW components in this room, the team :
discovered. a PVC line from the Polymer Addition System had '
. separated at a joint. This line was connected to the SWS by normally' ,
open manual isolation valves. Service water pressure was being held l by a non-safety related check valve in the piping. After the problem l was identi6ed, the two isolation valves were closed and DR N-94-982 :
was generated for the condition. During investigation of the ;
condition by the station, the PVC piping separated at another joint. ,
The response to Question #053.0 indicated the line had not been used ;
for one to two years and that the isolation valves would be changed .;
to normally closed in drawings and procedures. The design i implications have been discussed in Section 01.d.
i During an assessment walkdown, the team discovered high water in the SW Valve Pit in the Turbine Building. The problem was caused i by leakage back through the pit's sump pump check valve. The sump l pump could not perform its function because the discharge hose had >
separated from the pump. In addition, the level switch that should ;
have activated a Control Room alarm had not functioned. If the leakage had continued, water would have covered the SW to discharge tunnel MOVs, eventually filling the valve pit and leaking i over into the Auxiliary Building. The MOVs in the pit would be i manually aligned when establishing lake-to-lake operation, and do ;
not have an automatic function in response to an accident. DR N !
1004 was written for the condition and a work request was initiated to correct the problems (Ref. Question #092.0). j During the walkdown of the SW pump house, the four Service Water '
traveling screens were found to be severely corroded. Their condition is discussed later in this report in Section 03.f.
03.b. Witness maintenance on SW system. Haview WO package. !
Observe OC involvement. !
1 Preventative Maintenance (PM) activities conducted during the !
assessment were witnessed. These include removal of corrosion l monitoring coupons from Service Water piping in various locations and maintenance on the motor for one of the four SW pumps,1-SW- ;
P-1B. During the PM for removing corrosion coupons,it was noted i that a written procedure was not being used. The team was l concerned that a maintenance activity which involved opening and !
closing valves in the SW system was being performed without using a procedure. The justification for this was requested (Ref. Question
- 011.0), since it appeared this was not in accordance with usual station practice. The station agreed that such a procedure would be developed and a request was sent to the Supervisor of Station Procedures for the development of a new procedure for the replacement of corrosion coupons. .
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The package for maintenance on the motor for pump 1-SW-P-1B was !
reviewed and found to be adequate. Work Order #00287816 i implemented model work order E-15-M/C-2. for cleaning and i inspecting motors. 0-EPM-1412 01, Rev 2, " General Inspection and i Testing of Electrical Motors" was performed to measure phase to l phase resistances of the motor windings and Polarization Index (PI) ;
of the motor. The acceptance criteria for these parameters were l verified to be satisfactory. The work order implemented the !
requirements of VPAP-1302, " Foreign Material Exclusion i Requirements." After maintenance, a Post Maintenance Test (PMT) I was performed to verify operability of the pump and motor assembly. l The readings for motor current were verified to meet the acceptance '
criteria Section 7.2 of procedure 0 EPM-1412-01.
1 The team observed that the auxiliary SW pump discharge piping was j being painted and the missile shields had been removed to gain ;
access to the valve pit. The team inquired about controls to ensure '
the shields would be replaced prior to the occurrence of adverse weather conditions (Ref. Question #009.0). The response indicated !
that the missile shields would be replaced using 1-AP-42 (the i applicable procedure is actually 1-AP-41, " Severe Weather !
Conditions"). This was found to be acceptable.
l 03.c. Maintenance nrocedure technical content. Instructions to I identify deficiencies. consistancy with vendor ==nnale. I Complete and updated vendor manuals Vendor manuals for the Service Water Pumps, the SW Screen Wash Pumps, the SW Pump Motors, and the SW Pump Supply Breaker ,
Protection Relays were reviewed and compared to the maintenance I procedures for these components. Also the maintenance procedures '
used for inspecting and cleaning the CC HXs, Control Room chiller condensers, and the Charging Pump lube oil, gear, and seal oil j coolers were reviewed for adequacy. !
l The vendor manual for the SW and Auxiliary SW Pumps was found '
to contain generic information and did not give specifics for disassembly and reassembly of the pumps. The procedure ( 0-MCM-0115-01, Rev 1) used for performing maintenance on these pumps was found to be technically adequate. It is very detailed and includes instructions for setting lift, inspecting for erosion, corrosion and wear, coupling the various pump stages together, and setting running clearances for the internal components of the pump. It also includes instructions for Foreign Material Exclusion Requirements and QC Hold Point (s) at important steps. The team inquired about the basis for the clearances given in the procedure and requested a verification of their accuracy. No basis was documented and station personnel contacted the pump manufacturer who verified that the present clearances given in the procedure were correct. Docum6ntation was Page 47
l CORPORATE NUCLEAR SAFETY ASSESSMENT l NORTH ANNA POWER STATION SWSOPA l l obtained from the manufacturer stcting that the clearances i contained in the procedure are adequate. This documentation will be incorporated in the vendor manual for future reference (Ref. Question
- 054.0).
The procedure for performing maintenance on the SW Pump Motor was reviewed against the updated vendor manual,59-G533-00001, for adequacy and no problems were found. The instructions contained in procedure 0-EPM-1401-01, used for inspection and testing of the SW pump motor, were consistent with the requirements of the vendor manual.
Instruction manual L-64, "For Limitorque Type SMB Limitorque Operators," was reviewed for adequacy and no problems were found.
The requirements of the manual were implemented in procedure 0-ECM-1502-05," Inspection and Repair of Limitorque Valve Control Units, Type SMB-000 and SMB-00, with HBC Actuators for Butterfly Valves."
The updated vendor manual, 59 G533-00037, Type IAC Time Overcurrent Relays was reviewed and found to be technically adequate. It requires that as found data to be measured and recorded, checks for physical damage to the relay, checks for moving parts for obstruction, burnishing of relay contacts, and cleaning the relays. Procedure 1-EPM-1815-03, Rev 0, used for performing maintenance on breaker 15H4, contains instructions that are consistent with the requirements of the vendor's manual.
This manual also recommended an inspection frequency of six l months for the protection relay for the SW pump breakers. The :
present frequency of two years is required by the PM program. The team was concerned that relay maintenance was not being performed at the frequency recommended by the manufacturer. In response to the question on the adequacy of the maintenance frequency for the ,
relays, the Engineering responded that the present frequency is I adequate and provided justification from the manufacturer, General I Electric, to this effect (Ref. Question #109).
The procedure for PM on the Charging Pumps, 0-MPM-0103-01, inspects the SW side of coolers associated with the Charging Pumps.
l The assessment team identified that this procedure did not contain torque values for reassembly of these coolers (Ref. Question #105.0).
Station Procedures revised the procedure to specify torque values during the assessment.
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CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA 03.d. Maintenance corrective action for sittina. biofoulina.
corrosion. erosion and failure of nrotective confina.
Maintenance corrective action for silting, biofouling, corrosion, erosion, and failure of protective coating was reviewed and the following noted:
Sorav Arrays The spray arrays are periodically inspected to identify blocked nozzles and drains in accordance with 1-PT-75.11. Work Requests are then written for maintenance to clear the blockage. The cause of the blockage is not documented on completed Work Orders. The cause of the blockage could have been mud / silt accumulation or foreign material such as corrosion products, protective coating, or clam shells. Effective corrective action to prevent recurrence is dependent on identifying the type of material causing the blockage.
Comnonent Cooline and Control Room Chiller Heat Exchangers Inspections of the Component Cooling Heat exchangers (CCHXs),
Charging Pump coolers, and the are performed on a regular basis in accordance with the PM schedule to ensure that fouling of the heat transfer surface does not prevent the HXs from performing their design function. During several of these inspections, mud and silt was observed to have accumulated in the CR chiller condensers, CCHXs, and charging pump coolers. In all of these instances the HXs were considered capable of removing the design heat load.
These inspections do not completely document the as found condition of these HXs. For instance, procedure MMP-P-R-1, " Periodic Disassembly, Inspection of the Control Room Air Conditioning Chillers," Step 7.5.5 states, " Chemistry shall visually inspect the as-found condition of the service water side of condenser. This inspection is for evidence of Asiatic clams (corbicula) shell debris, or any foreign material which could cause flow blockage." It does not require documenting the actual as found condition (Ref. Question
- 048.0). Also, step 6.2.7 of procedure 0-MPM-0103-01 only requires recording SAT /UNSAT for indications of erosion, corrosion, foreign material, and mud in the charging pump seal, lube oil, and gear coolers. There is no requirement for documenting the conditions found, the magnitude of these parameters, or any criteria for determining if the condition is SAT or UNSAT. Action III of GL 89-13 indicates, "results of these maintenance inspections should be documented." Procedures used to perform inspections of HXs for the purpose of GL 8913 compliance should be revised to l require appropriate documentation of the as found conditions (Ref. Question #048.1). .
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I CORPORATE NUCLEAR SAFETY ASSESSMENT l I NORTH ANNA POWER STATION SWSOPA l .
Recirculation Sorav Heat Exchanaers A review of the history for the RSHXs has shown that inspections of the internal conditions were performed in 1988 after it was identified i that the RSHXs needed to be maintained in dry layup (Reported in i Unit 1 LER 88-16). No additionalinspections of the RSHXs have i been performed or are planned. Each refueling outage these HXs are exposed to SW for a few hours during SW flow balance and CDA '
functional test proceduros. At end of these tests, these HXs are ,
drained via procedure 1/2-MOP-49.3. However, no physical ,
inspection of the interior of the HXs has been performed to validate
- the effectiveness of this procedure to ensure that mud and/or silt ,
have not accumulated on the heat transfer surfaces of the tubes. ;
Inspection of the CCHXs, charging pump coolers, and control room ,
chiller condensers have shown that mud and silt can enter the tubes ;
of heat exchangers supplied with SW, although the time period the -
RSHXs are exposed to SW is much less than these other HXs.
The team asked the question about the effectiveness of procedure 1/2- I MOP-49.31 and the frequency at which inspections are planned for the RSHXs. The team was informed that an inspection had not been planned and that one will be considered for the future. The concern about the effectiveness of the inspection procedure was not i addressed. The RSHXs should be visually inspected ;
periodically to verify their cleanHuess (Ref. Question #119.0). j Intaka Structures The intake structure, along with the suction bells of the auxiliary SW
- pumps and SW pump screens, are inspected for corrosion, erosion, silting and biofouling. In 1990 inspection of 2-SW-P-4 and 2 CW-P.
IA by a diver indicated "no great accumulation of[ asiatic) clams and buildup on floor and intake screens." -
Pinina
, The response to Generic Letter 89-13 dated January 29,1990 stated,
- " North Anna Power Station does not have a history of biofouling causing flow blockage in the service water system. Accordingly, periodic surveillance of representative service water components will l demonstrate the continued lack of biofouling." Technical Report No. 1 ME-0025, Rev.1 also indicates that biofouling has not caused flow blockage in the SWS. The Technical Report also states, "The existing chemical treatment program is sufficient to control biological !
macrofouling and the gross deposition of corrosion products that I could significantly affect flow balance in the Service Water System." ;
Although mud and silt have been found in the CCHXs, control room l chiller condensers, and charging pump coolers during inspections, l
. flow blockage in SW piping has not been a problem. The primary ;
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CORPORATE NUCLEAR SAFETY ASSESSMENT I
NORTH ANNA POWER STATION SWSOPA ,
concern with the piping is degr'adation caused by Microbiologically - ;
Influenced Corrosion (MIC).
Service Water Preservation Proiect !
The station implemented the Service Water Preservation Project to (1) address and attempt to eliminate the root cause of the corrosion ;
damage to the SW system piping, (2) prolong the remaining life of the i currently acceptable portions of the system piping, and (3) provide '
recommended conceptual repair and/or replacement designs for heavily damaged sections on a " remaining code life" priority.
This project is divided into two parts, designated as Phase I and 'I Phase II. The Phase I activity addresses refurbishment of the buried and concrete-encased 24" Main Service Water headers which '
included the supply and return piping to the Unit 1 Recirculation
. Spray Heat Exchangers (RSHXs), the supply and return piping to the i Unit 2 RSHXs, the Auxiliary SW supply and return piping, and the ,
supply and return piping to the Units 1 and 2 component cooling heat exchangers (CCHXs). Phase II will include 36" piping coating ,
inspection and refurbishment, SW Reservoir cleaning and lining,- ,
refurbishment of the remaining portions of the exposed SW piping, and total system cleaning. ;
To determine the starting dates for Phase II, condition monitoring of ,
the SW piping is performed using Periodic Test 1-PT-75.14. This :
procedure performs UT to document the corrosion rate and remaining wall thickness of the SW piping at various predetermined locations.
The data obtained from this PT is reviewed by Engineering and Engineering Mechanics to accurately trend corrosion rates and ,
compare the data against the requirements of NRC Generic Letter >
90-05 to assess the need for repair and to establish priorities for the i SW preservation project. l In addition to performing PT-75.14, on-line corrosion monitoring of !
the SW piping for different flow conditions (stagnant, intermittent or +
continuous) is performed to assess system corrosion and the effectiveness of the chemical treatment program. Also an extensive l SW chemical treatment program is in place to inhibit corrosion and control biological growth and bacteria attack.
Since the inception of SW preservation project over 1000 ft of 24 inch SW piping and 750 feet of the Auxiliary Service Water Pump discharge piping has been replaced or repaired with coated piping.
The implamentation of the Service Water Preservation Project is considered a strength.
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CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA Chemical Control Program Aspects of the SW chemical control program were reviewed to determine its technical adequacy, implementation and effectiveness.
The elements of the program are as follows: 1) a bromine-based biocide for algae and aerobic bacteria 2) a molybdate based cor osion inhibitor 3) a polymer to be used if suspended solids exceed limits 4) two biocides used for bacteria and MIC control as needed.
Based on a review of the technical evaluations of the program, the general objectives of controlling biological growth and minimizing corrosion by the chemical control program appear to be accomplished.
There are some further optimizations to be made, such as installing a larger brominator to increase the feed rate of the bromine based biocide. There is also an indication that the Total Suspended Solids (TSS) are low. This is important to the NAPS SW system due to some instances where the chemical treatment and/or rain have caused some of the clay liner material to be suspended in the reservoir. The entrained liner material has then fouled plant equipment. Recent TSS data indicate this problem is being controlled.
During the review of the chemical control program, " Service Water Chemistry Log" chemical control parameters and SW system chemistry data were examined. Acceptance criteria were not on the log sheet, so Section 5.2.12 of the Nuclear Plant Chemistry Manual was reviewed to determine operating limits. Only one specification was established for the NAPS SWS. This was a 2550 ppm limit for TRC-256, the corrosion inhibitor.
i The 2550 ppm limit was compared to operating data since 12/16/93. !
Five instances where TRC-256 was below the specification were ;
fout.d. Reasons for the low values were not noted in the log. In four i of the five cases, the corrective action for the low readings was not i documented. Subsequent in-specification readings in each case indicate corrective action was taken. In the fifth case, a low reading of 468 ppm had a remark 'of " suspect bad sample" but the next reading (which was within specification ) was not recorded until two days later.
The team questioned the timeliness and documentation of the corrective actions for these parameters. There is no technical problem with TRC-256 chemical control because of the conservatism in the specification and because sustained operation at low values would be required before corrosion of the SWS would be significantly affected. However, corrective action for an operating parameter on a safety related system should be timely and clearly documented, l espec; illy on a system with a history of problems related to corrosion.
The ntation committed to taking resamples in a more timely rnanner i
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CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA
'l E I and to more clearly document the corrective action (Ref. Questions 003.0 and 003.1). This satisfies' the team's concerns in this area.
One other question concerning SW system operation was identified during the chemistry review (Ref. Question 016.0). Based on Calgon's (the chemical vendor) silica readings _of 7.0 and 7.4 in lake !
water, and SW system silica readings that range between 28 and 31. !
ppm, it appeared the SW system was running consistently at- j approximately 4 cycles of concentration. It had been stated in a training session that SW system blowdown is not used, so it was not ,
apparent how dissolved solids were being controlled and what effect the build up of dissolved solids might have after operation for the '
design basis of 30 days without make-up.
The response to these questions indicated dissolved solids were not .
controlled and were not a problem due to the low solids content of the make-up water from Lake Anna. It was also later confirmed that the bearing cooling water system cooling tower also does not operate a blowdown, apparently depending on drift to remove dissolved solids. l The final concentration of solids in the SWS should be roughly double i the initial value at the end of the 30 days. Given this information, it ;
is agreed that the dissolved solids should have little impact on system ;
operation during normal or accident conditions. i 03.e. Adequacy of maintenance for accident conditions. Raylew !
unavailability due to nlanned maintenance l
Review of the maintenance history indicates that, in general, maintenance and testing of SWS components ensure that equipment i will function during accident conditions. Testing of the components i is discussed in Section 04 of this report. Preventative maintenance )'
(PM)is performed to preclude degradation of equipment and ensure equipment operability during accident conditions. Review of the PM ;
and Reliability Centered Maintenance (RCM) program indicates that a comprehensive maintenance program has been implemented for the ,
SWS. The scope of the program includes system heat exchangers, ;
check valves, relief valves, pumps, motor operated valves, motors, breakers, relays, and instrumentation. The team found PM scope 1
- and frequency on some SW equipment may not ensure the long term ,
reliability of those components. These issues are discussed in more :
detail below. ;
Auriliary and Main Service Water Pumns Review of the maintenance history for the screen wash pump,1-SW-P-2, indicated that in 1992, inspection of this pump indicated i excessive wear, scoring and distortion of the pump components such ,
as tho shaft sleeve, head bearing, and line bearings for each of the :
pump stages. Also, deterioration of the pump protective coating on i i Page 53 ,
CORPORATE NUCLEAR SAFETY ASSESSMEN'1 NORTH ANNA POWER STATION SWSOPA the internal and external surfaces was noted. The team asked what inspection is planned for the SW pumps to ensure that similar degradation will not occur (Ref. Question #010.0). The response stated that no inspections are planned and pulling SW pumps for inspection should be performance based and that the exterior of one pump would inspected during maintenance of the associated traveling screen at the refueling outage. It was further stated this would give an accurate indication as to the condition of the coatings on the remaining pumps. Due to the significant differences in velocity and flow between the interior and exterior of the pump, the team considers that such an inspection will not give a true indication of the internal condition of a SW pump.
The RCM program recommends that preventative maintenance l should be performed on the Auxiliary and main SW pumps on at least a ten year interval. This PM should inspect and/or refurbish ,
pump internals such as pump bowls, pump impellers, line bearings, l shaft etc. Maintenance was last performed on the main SW pumps ,
between 1982 and 1983. It has been almost eleven years since this !
maintenance was performed. Operatior.al test data has indicated ;
that the pump performance curve has degraded below its initial value '
even though each pump is presently capable of delivering the required flow during accident conditions. Operational test data does :
not give an indication of the condition of the pump internals such as l pump column corrosion and pump bowl and impeller )
erosion / corrosion. The team recommends that the PMs for l disassembly, inspection, and refurbishment for each of the i main and auxiliary SW pumps be planned and performed in accordance with the requirements of the RCM Program (Ref. l Question #010.1) )
l Heat Exchancers for the Instrument Air Comoressors The two station Instrument Air (IA) Compressors each have three heat exchangers which are cooled by SW. In late 1993/early 1994, !
the RCM program recommended that the SW side of these heat exchangers be cleaned every three years. The station indicated that procedures are being developed for performing this work. Although the IA Compressors are not safety related, IA is required for normal station operation and is useful in operating the plant during j transients and accidents. Since the heat exchangers are not isolate during a DBA, they are part of the SWS pressure boundary as well.
The team agrees the recommendations of the RCM should be implemented (Ref. Question #111).
Service Water Pumn Motors The PM program ensures that the SW pump motors are capable of delivering the required driving torque to the SW pumii s during the i
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design basis accident by performing preventative maintenance (PM) 1 on the main and auxiliary SW pump motors once per quarter. This quarterly PM checks for freedom of rotation, internal rubbing, binding, or noise as applicable, phase-to-phase resistance of the I motor windings, and the Polarization Index. Also, PM is being l performed on the main and auxiliary SW pump motors every five -
years to disassemble, clean, and/or refurbished them as necessary in accordance with the RCM program. I In 1988, in response to IEIN 87-30," Cracking of Surge Ring Brackets in Large General Electric Company Electric Motors," the motors were inspected and cracking of the end winding was discovered. Repairs '
were made to the windings during this inspection. A review of the maintenance history did not initially identify follow-up inspection of these motors and the team inquired (Ref. Question #114.0) what is the justification for not performing the recommended maintenance.
In response to this question the maintenance department provided documentation which showed that the motors of both the main and auxiliary SW pumps had been disassembled and inspected in 1992 as recommended. The inspections revealed no damage to the surge rings and validated the effectiveness of the 1988 repairs.
SW Pnmn Traveling Screens Preventative maintenance is being performed on the SW pump screens. However, this maintenance has not corrected the degradation of the screens. This is further discussed in Section 03.f Ernansion Joints Review of the maintenance history for the SW system expansion joints indicated that they had not been in a PM program prior to their replacement in 1989. Discussions with maintenance personnel and review of the RCM program indicated that there was no planned frequency for joint replacement and that this would only take place after inspection had indicated the need for replacement. The team inquired as to the basis for the present replacement frequency. The response indicated that the vendor's marnal showed that the shelf and service life of the expansion joints is five years under ideal storage conditionc. The present inspection frequency is 3 years and l no PM has been established for replacing the expansion joints at any frequency. The teault is that expansion joints may not be replaced within the expected service life. A DR was submitted (N-94-992) to identify and evaluate the discrepancy between the present PM and the vendor manual (Ref. Question #050.0).
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CORPORATE NUCLEAR SAFETY ASSESSMENT I NORTH ANNA POWER STATION SWSOPA
' 03.f. heurrina caninment nrohlema and trends. Root Cause !
annivain and corrective action for adverse trends. Work nacirman technical admananv. PMT. and demonstration of operabilitv.
The maintenance work history from 1991 to the present was reviewed. Components with recurring problems included the RSHX -
relief valves and the CCHXs.
RSHXs Tube Side Relief Valves The SWS relief valves are being maintained in accordance with the :
safety and relief valve program. Several work orders for '
maintenance on the RSHXs relief valves and SW air compressors were reviewed for recurring problems and trends. Two work orders for the Unit #2 RSHXs are fully discussed to indicate the scope of this review and typical problems found with the relief valves.
- WO #00259613 " Perform removal, testing, repair, and installation of safety relief valve 59-02-SW-RV-200B in accordance with 0-MCM-0400-30 and 2-PT-147.1." The valve lifted at 145 psig which was below the setpoint of 146 psig and was rebuilt under WO #259613-02. The work order included Foreign Material Exclusion (FME) requirements in accordance with VPAP-1302 ,
and testing was satisfactorily performed. The procedure 0-MCM-0400-30 was reviewed and was found to be technically adequate and included requirements for documenting the as found leakage
and lift setpoint. :
- WO #00259612-02 " Perform removal, testing, repair, and installation of safety relief valve 59-02-SW-RV-200A in accordance with 0-MCM-0400-30 and 2-PT-147.1." This valve failed testing and was disassembled, inspected, and cleaned. It was noted that the valve was " inspected and cleaned all parts and cleaned all rust and trash out of valve. Found trash on seat and disc. Lapped seat and disc and rebuilt valve." The valve was then satisfactory ,
retested in accordance with 2-PT-147.1. In 1990 this relief valve failed to operate as 6? signed as recorded in DR 90-1515. The probable cause was determined to be " Debris in the system".
The maintenance history for the Unit 1 RSHXs relief valves were also reviewed and similar problems noted. The setpoint for relief valve 59-01-SW-RV-1000 was found to be outside the allowed setpoint range of 146 psig to 154 psig. This is documented in DR-N 91-296.
Review of the maintenance history of the SW relief valves indicates that the relief valves for the tube side of the RSHXs have consistently failed to satisfy the requirements for seat leakage and lift setpoints.
This has resulted in the replacement of the failed relief valves. In all Page 56
CORPORATE NUCLEAR SAFETY ASSESSMENT l
NORTH ANNA POWER STATION SWSOPA instances it was determined that failure was due to debris in the system. Over the years SW flow through the RSHXs has resulted in the formation of corrosion products in the discharge piping to the relief valves. These corrosion products may have been deposited in the seating surface of the relief valves during the periods that SW was flowing through the HXs. Coating of the internals of these discharge lines with corrosion products can result in the lines becoming partially restricted. It appears the root cause for relief valve failure is not being addressed. It is recommended that piping to each relief valve be inspected and/ or cleaned. The need to inspect and/or flush these lines after they are exposed to SW flow prior to reinstallation of relief valves should be evaluated (Ref. Question #152.0).
Service Water Air Comnressor Relief Valve The maintenance history of the SW air compressor relief valves were reviewed. Two work orders are discussed here: .
- WO #002289820 01 - This non-safety related work order was performed on 06-16-94 to replace corroded relief valve 01-SW-RV-111A, air compressor 1-SW-C-1A discharge relief valve.
Procedure 0-MCM-0400-31 which was used to perform the work on this relief valve was reviewed and found to be technically adequate. It required documentation of lift pressure and seat leakage and the PMT performed was adequate to verify operability of the relief valve.
- WO #00262514 This work order was performed on 08-26-93 to replace air compressor,01-SW-C-1B relief valve,01-SW-RV-111B.
Since the relief valve was deemed not to be adjustable, the decision was to not test it prior to installation. This assumption was incorrect since the relief valve is similar to 01-SW-RV-111A -
and can be adjusted as demonstrated in WO #002289820-01. Both valves are manufactured by Kunkle and their suitability for use in this application was determined by Item Equivalency Evaluation Report N 93-5025.000.
Service Water System Travelina Water Screens Review of the maintenance history of the traveling water screens indicated that since 1989 the screens were determined to be corroded and required replacement. A walkdown during the assessment ,
verified they were stillin this condition. Several work orders were reviewed and included:
WO #5900154685, completed on 01-07-93, performed procedure M-10 TS/SA on SW traveling water screen,01-SW-S-1A. It was noted on this work order that the screens should be overhauled and WO Page 57
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA
- 095693 had previously been submitted. (This WO, originated in 1989, had not been started at the time of the assessment.) Over one year later, on 4-28-94, step 6.8.1 of PT-75.4 was signed off that erosion and corrosion of the screen was satisfactory in spite of the degraded condition of the screen. It appears personnel responsible for performing the procedure did not fully understand how much corrosion was acceptable. The procedure did not provide direction in this area (Ref. Questions #076.0 & 077.0).
Work Orders #00287818-01 and #5900161216-01 were also used for maintenance on the SW traveling water screens. The two work orders were both signed off as completed satisfactorily along with PT-75.4, which verified the operability of the screens. In spite of the fact that it was noted that the four SW traveling water screens are severely corroded and should be replaced or removed from the system, no DR was written between 1989 and June 1994 to document this deficiency. The team expressed concerns (Ref. Question #012.0) about the structural integrity of the screens. In response to this concern DR N-94-0997 was issued and Safety Evaluation 94-SE-OT-046 determined that the current condition screens would not adversely affect the operation of the SW system. (The team has made a recommendation in Section 02 about the need for these screens.) In addition, PT 75.4 should be revised to include clear acceptance criteria for the inspection of the screens for erosion and corrosion (Ref. Question #070.0).
SW system check valves A maintenance program which includes performance testing and visual inspection has been established for check valves in general. l Review of the maintenance history over the last two years indicates that the program has been effective. The following work orders are discussed:
- WO #5900152458. This work was used for maintenance on check valve 1-SW-343 using procedure 0-MCM-0463-01, "Vogt Piston Check Valve Repair." The procedure is technically adequate. It provides detailed instructions for documenting as found and as left conditions of the valve internals using Attachment 4, Maintenance Data Sheet, and includes QC hold point. Post ;
maintenance testing was performed to verify valve operability.
- WO #5900154463. This work order performed maintenance on auxiliary SW pump discharge check valve,1-SW-22, using procedure 0-MCM-0431-01, " Mission Duo-Check and C&S Dual Plate Check Valve Repair." The procedure is technically adequate and provides detailed guidance for performing as found and as left inspection of the valve components for signs of degradation from crosion and corrosion. Instructions is also included ~for en'suring l
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that the proper torque values are used during valve assembly and that a close-out inspection is performed in accordance with VPAP-1302. In addition, detailed PMT instructions were provided to -
ensure that post maintenance testing was technically adequate to verify proper valve operation when opening and closing. ,
WO # 00275159-01. This work order was initiated to remove and reinstall check valve 01-SW-1125. The valve had been previously i installed backwards. DR N 93-1784 was written to document the faulty installation. It was determined that the valve was- ,
improperly installed because a procedure was not used to perform i the work. To prevent recurrence management decided that future work on similar check valves would be performed using a !
procedure instead of by skill of the craft. A CTS item was then i assigned to the procedures department to add a step to any i generic check valve procedure to verify proper direction of flow on :
reassembly. The team verified that procedures,0 MCM-0400-12, i
" Disassembly, Inspection, and Repair of Non-Safety Related i Check Valves," and 0-MCM-400-32, " Disassembly, Inspection, and ,
Reassembly of Safety-Related Check Valves in General," had been i developed for performing future work on check valves. These procedures contain guidance for verifying valve operability after maintenance. ;
- WO #5900152463. This work order was initiated to disassemble, inspect, repair, reassemble, and test the screen wash pump ,
discharge check valve,1-SW-63. Procedure 0-MCM-437-01, i
" Pacific Swing Check Valve Repair,"is technically adequate and provides detailed guidance for performing as found and as left 1 inspection of the valve components for signs ofleakage, damage to the valve internals, fasteners, hinge pin plugs, and plug holes. It
- also provides instruction for reassembly, gives specific values for ;
torque, and requires QC verification at various hold points. Post -
l maintenance testing was performed to verify no external valve :
l leakage in accordance with procedure PMP-LKT-MM-0001.
03.g. Maintenance personnel technical knowledge of component maintenance. settina limit switches. numn counlina alianment. cleanina/renlacina filters. circuit brealrar maintenance. l During observation of maintenance activities, maintenance personnel demonstrated satisfactory technical knowledge needed to perform their assigned tasks. Knowledge is also demonstrated during performance of Job Performance Measures (JPMs) during developmental training.
Interviews were performed with plant personnel to determine how components were maintained. The interviews revealed that Page 59
CORPORATE NUCLEAR SAFETY ASSESSMENT I NORTH ANNA POWER STATION SWSOPA individuals performing maintenimee on SW pumps, valves, piping, SW screens, and HXs, etc. are adequately trained and their technical knowledge in these areas were more than adequate to ensure that maintenance performed on SW components would be adequate to ensure the continued reliability of the SW system.
03.h. Trainina of maintenance nersonnel and consistency with nrocedures.
i Specialized training is not provided for inspections of SW system components. Personnel are given training through their department ,
development program (Ref. Questions #048.0 & #048.1). Procedures require Chemistry personnel to perform inspections for silting, biofouling, corrosion and erosion. In response to questions by the team, procedures will be revised so that future inspections for silting and biofouling will be performed by either System Engineering or Maintenance Engineering. This does not address the training '
qualification requirements for inspections or provide guidance for documentation of as found conditions, acceptance criteria, and l threshold for deviation reports. It is recommended that special l training be provided for personnel who perform inspections I of components covered by GL 8913. Topics should include GL .
8913 requirements and commitments, recognition of Asiatic ;
clams / zebra mussels, and expectations for documenting as i found conditions (Ref. Question #048.1).
The training received by the maintenance personnel for performing maintenance work on the SW system is adequate and is consistent with the amount of technical detail in the procedures reviewed.
Maintenance personnel are required to complete JPMs on flanged joints, instrumentation such flow indicators and transmitters, vertical turbine pumps, rotary compressors, Limitorque operators, relief and safety valves, control valves etc. In addition maintenance personnel undergo initial and continuing training in several areas which includes proper interpretation of plant documentation, root cause analysis, plant chemistry, computer training, breaker O & M, Raychem Splices, SOV maintenance and mechanics of MOVs.
03.i. Insnection nrogram to detect corrosion. erosion. nrotective conting failure. siltina. and biofoulina.
The inspection program is described in VPAP-0811, " Service Water Inspection and Maintenance Program." This procedure lists the major components to be inspected and lists many activities which are to be followed to enhance the implementation of GL 8913.
Numerous station procedures implement inspections of SW components. Periodic Test,0 PT 75.15," Generic Letter 8913 Service Water System Testing Requirements Coordination," is performed annually to verify that inspections, evaluations, and proc'edures have Page 60
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been performed to meet GL 89-13 requirements. The PT verifies that the following have been performed:
Over forty other periodic tests which perform inspections of SW components PM procedures which inspect and clean the Charging Pump coolers and coolers for the Control Room chillers Diver inspection of the intake structures Periodic Test 0 PT-75.15 also documents the results of the clam survey report and the annual corrosion report.
Ultrasonic Testing (UT) of wall thickness of uncoated SW piping is performed in accordance with 0 PT-75.14. The procedure calls for examination of 47 points in various piping sections. The results of i the UT are reviewed by Corporate Engineering to determine wall thickness (general and under pits), determine corrosion rates, and ensure that minimum wall thickness will be maintained until the next outage. The results of the evaluation are documented on Calculation NE-0149. This report is attached to 0-PT-75.15.
The team identified that not all of the instructions contained in i VPAP-0811 are being followed. For instance, the procedure calls for a dedicated SW file for " Reports documenting the basis, and subsequent changes, for selecting the methods for meeting Service Water Inspection and Maintenance Program requirements" be maintained in Station records. Such a file is not being maintained.
The VPAP requires the type and quantity of fouling be documented as part of the inspection results. Inspection procedures for the Charging Pump coolers, the coolers for the Control Room chillers, and the CCHXs did not require that these inspection results be documented. The results were not documented on the Work Orders associated with these inspection procedures (Ref. Question #048.1).
Step 6.1.3.g of the procedure states "Where feasible, visual records of conditions observed should be recorded by photographic means." '
Photographing of as found conditions is not being performed and implementing procedures do not provide directions for photographs.
Photographs could provide trending information relative to fouling which could be used to change inspection or cleaning frequency.
It is recommended that the station comply with VPAP-0811 >
(Ref. Question #049.0).
Service Water Pumo / Coatinc Insoection The team also noted that the following is not being done as part of the station PM program or VPAP-0811:
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CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA Inspection of the SW pumps including the pump line bearings, pump bowls, and pump columns (Ref. Question #010.1)
Inspection of the pump bowls and pump column coating for degradation.
Inspection of the coating of SW piping where applicable. (Ref.
Question #055.0)
Applying a protective coating onto SW piping is being performed as part of the Service Water Preservation Project. Some piping has been recently coated and additional piping will be coated when the pipe is replaced or repaired. When asked about plans for inspecting the new pipe coating, Engineering indicated that informal inspections of recently applied coatings indicated that the coatings were in excellent condition. However there was no formal inspection program and "The need for a formal inspection program at North Anna following completion of the project will be evaluated at a later date." Item III of GL 89-13 recommends establishing a routine inspection and maintenance program to detect protective coating failure. The GL further states "A description of the program and the results of these maintenance inspections should be documented. All relevant documentation should be retained in appropriate plant records."
Protective coating has been applied to SW pumps. As discussed in section 03.e, deterioration of protective coating on SW screen wash pump,1-SW-P-2, was noted in 1992.
Recommendations for inspecting Service Water system pumps and pipe coating have been made in other sections of this report.
To prevent corrosion of SW piping, cathodic protection was installed by DCP 84-107-3 which was approved for installation in 1987.
Preventive Maintenance procedures were recently developed for maintaining cathodic protection system but have yet to be implemented. According to station personnel, portions of this cathodic protection system have not been placed in service. The cathodic protection program is not included in either VPAP-0811 or 0-PT-75.15. Since this system was installed, the Service Water Preservation Project has installed protective coating on some SW piping. This protective coating may require changes in the operation of the cathodic protection system, specifically in the location of electrodes and the current density applied. It is recommended that the operation of the cathodic protection system be reviewed and (f.ie effect of any new protective coating on affected SW piping on cathodic protection design be considered (Ref. Question #090.0).
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CORPORATE NUCLEAR SAFETY ASSESSMENT I- NORTH ANNA POWER STATION SWSOPA
- 04. SURVEILLANCE and TESTING A. SIGNIFICANT ISSUES l 1, Check Valve Testing !
Testing SW chemical injection system check valves '
PMT on Auxiliary SW pump check valve
- 2. Testing MOVs Under Worst Case Conditions ;
B. ASSESSMENT TOPIC DISCUSSION Review and Anoroach i In the area of surveillance and testing, the team reviewed a wide range of !
documentation related to ASME Section XI surveillance , testing, post- i maintenance and post-modification testing, and heat exchanger performance testing. Portions of the UFSAR, Technical Specifications, and i surveillance testing program documents, including relief requests, were :
reviewed to identify regulatory commitments. Design documentation such as design drawings, the system design basis document, performance testing ,
calculational basis, and design calculations and analysis were reviewed to i identify the system design and testing parameters. Finally, the team reviewed surveillance testing procedures including completed test results, surveillance testing trend data, work orders, and design modifications. The team also witnessed surveillance testing on the chemical addition >
subsystem boundary manual and check valves. :
The purpose of these reviews was to assess whether the surveillance and ,
testing program at North Anna provides adequate assurance that the SWS will perform as designed. Generally, the team found that the surveillance l and testing program provides adequate assurance that the SWS will :
perform as intended. Also, the team found that the design and testing '
documentation was generally retrievable and the engineering and testing !
personnel were committed to improving the overall testing program and the ;
quality of the work they perform. '
04.a Technical adequacy and accuracy of TS surveillance and IST I procedures. Incorporation of design assumptions. t The Technical Specification (TS) surveillance requirements were ,
reviewed to determine whether they adequately demonstrate component and system requirements, including design assumptions. i No concerns were identified. The following is a list of the TS surveillances and the results of the review performed:
I Page 63
. _ __.. _ - - .,._.__-m. . ._-. .,y. ,..y.. ,.- , ,
CORPORATE NUCLEAR SAFETY ASSESSMENT I NORTH ANNA POWER STATION SWSOPA 1 Seismic monitonng instrumentation for the Component Cooling Heat Exchanger (CCHX) Triarial Peak Accelerometers. Channel :
Calibration is performed every refueling as required by the TS per procedure 0-PT-39.4.
Triaxial Response-Spectrum Recorders for the CCHX Support. A channel check requirement exists to visually inspect to detect signs of obvious physical damage. A Channel Calibration is performed at every refueling per Procedures 0-PT-39.5 and 0-PT-39.6.
Verifying at least once per 31 days that each valve servicing ;
safety related equipment that is not locked, scaled or otherwise l secured in position is in its correct position is performed by '
procedure 0-PT-75.1. Although the team determined that the procedure is adequate, in several instances, it was difficult to determine the actual position of the butterfly valves.
Measuring the movement of the SW pump house and wing walls at least once per 6 months. This is performed by procedures 0 PT- ,
115, i
At least once per 18 months during shutdown verify that each ,
automatic valve servicing safety related equipment actuates to its i correct position on an actual or simulated safety injection signal.
This testing is performed by procedure 1-PT-57.4.
- Verifying that each automatic SW valve actuates to its correct position on an actual or containment high high signal. This testing is performed by procedure 1 PT-66.3.
Each SW pump shall be tested in accordance with Specification l 4.0.5. This testing is performed by procedures 1/2-PT-75.2A.1.
The ultimate heat sinks, the SW reservoir and the North Anna )
reservoir shall be demonstrated operable by verifying at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that the average water temperature and water level are within their limits. This is performed in accordance with 1- i LOG-4 (SW Reservoir) and 1-LOG-6A (Lake Level). !
Data for calculating the leakage from the SW reservoir shall be obtained and recorded at least once per 6 months. See procedure 1 PT-75.8.
The total settlement of each Class I structure or the differential settlement between Class I structures shall be determined by measurement and calculation at least once per 6 months per procedure 0-PT-115.
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l CORPORATE NUCLEAR SAFETY ASSESSMENT l l NORTH ANNA POWER STATION SWSOPA l At least once per 6 months verify that the groundwater level within the dike of the SW reservoir does not exceed the value stated in TS Table 3.7-6. Measurements are for the SW pump house, south east end of reservoir, SW valve house. This is performed by procedure 0-PT-75.7.
Verify that the groundwater flow rate does not exceed the values established in TS Table 3.7-6. Also a visual inspection of the clarity of outflow from each drain shall be performed in conjunction with the flow monitoring effort. This is performed by procedure 0 PT-75.6.2.
Also reviewed were IST procedures to identify whether they were adequate to demonstrate system and component adequacy and whether design assumptions were incorporated into the procedures.
The IST procedures were generally found to provide adequate assurance that the pumps and valves in the IST program performed to the design requirements.
04.b SW design and licensing basis consistency with test acceptance criteria.
The team's review found that pump and valve testing was generally consistent with the design and licensing basis.
Pumo Testine The pumps in the IST program are tested to parameters that ,
encompass the design requirements as demonstrated by the l performance testing. For example: !
Service Water Pumos - Due to system operating conditions and heat removal requi:ements, IST testing performed on the SW pumps is based not on one operating point but rather on a band of acceptable points. Therefore, depending on the system heat load removal requirements, the flow is allowed to vary between 7,000 and 12,500 gpm and likewise, the corresponding pump developed head varies within an acceptable performance range established from the ASME Section XI upper and lower performance limits for the pumps.
i The IST testing range for the SW pumps is well within the design I requirements for the system even though the pumps are typically l not tested at the UFSAR value of a nominal 11,500 gpm (see l UFSAR section 9.2.1.2.1). However, testing to ensure capability I to perform to design requirements and to detect pump l degradation is performed at every refueling or after maintenance l affecting pump performance. The SW Pump Head Verification i test consists of developing a system head curve to determir.e Page 65 l l
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA whether the pump is performing above an " acceptable curve" which was developed from a reference curve degraded by 3%. The reference curve was established by field testing in 1989. If the pump curve falls below the " acceptable curve," an engineering analysis is performed to ensure that the system design conditions including the flow balance are not affected. (Ref. Question #103.0)
Auxiliary Service Water Pumns - These pumps are IST tested in accordance with the requirements of Section XI similar to the testing performed on the main SW pumps. As was the case with the SW pumps, IST testing does not necessarily simulate the design flow through the system. However, unlike the main SW pumps, no head curve verification procedure exists which would determine the ability to operate at design basis conditions (Ref.
Question #107.0). The Auxiliary SW system is a backup system and there is no specific requirement that a full head curve test be performed on these pumps. Also, since these pumps operate intermittently, it is expected that pump degradation will be minimal and should be detected during the quarterly performance testing of these pumps.
Check Valve Testine With respect to IST check valve testing the following was examined:
The quarterly IST testing performed on the SW pump discharge check valves was examined. The reverse flow testing of these valves uses as an acceptance criteria the check valve's inlet pressure of 10 psig or less. A pressure of 10 psig, on the inlet of the valve is indicative of gross failure of the valve because the flow path is to the open SW reservoir, although a review of the actual IST test results determined that the pressures recorded from the inlet side of the check valves were well below the acceptance criteria. These check valves had previously been tested (for approximately 6 months) by closing the manual isolation valve on the discharge of the non-operating SW pump and reading SW header flow. This manual valve was then opened and the header flow was then noted. The difference between the two flows was then the flow passing through the check valve, a quantitative indication of check valve leakage. This practice was changed due to an incident where during a SW pump discharge check valve test, a SW pump was operated at pump shutoff due to human error.
Although the previous test method can be seen as a "better" test, the quarterly check for gross leakage is in accordance with the specific request of the NRC in response to Revision 6 of North Anna's IST program submittal. The 18 month testing frequency for the check valves themselves is also approved as part of the program. Leakage through these valves in the reverse direction is checked, by inference, Page 66
i CORPORATE NUCLEAR SAFETY ASSESSMENT l NORTH ANNA POWER STATION SWSOPA during the quarterly SW pump by checking header flow vs. pump pressure on the head curve. Divergence outside of acceptance criteria will result in investigation, which will include consideration of a leaking check valve on the other pump on that header (Ref. Question
- 032.0).
Although it was ultimately determined there was no concern with ,
compliance with the IST program with this question, it was found that the acceptance criteria for backleakage through these check valves was not an input to the hydraulic model. This is mentioned in Section 01.e, Topic: " Flow Balancing "
MOV Testing The stroke time testing parameters for the MOVs within the IST program have in every case been determined to be within the system design requirements. Furthermore, the stroke timing for these valves is based on the valve operating characteristics which envelope the design parameters.
The GL 89-10 scope (MOVs) was reviewed and found to generally be adequate. However, the testing performed in the field may not be representative of the worst case design basis scenario.
During the review of the MOV differential pressure testing performed during the Service Water Flow Balance, by procedure 2-PT-75.6, between the dates 10-15 93 and 10-20-93, the team identified instances where the attained differential pressures (dP) across the MOVs were higher than the acceptance criteria provided in the procedure. Specifically the MOVs under question were the SW-X01s '
and SW-X03s. As indicated in the response to Question 121.0, the acceptance criteria was based on a calculated dP for post CDA scenarios and not the initial CDA sequence including consideration for the dry lay-up condition of the Recirculation Spray Heat 1 Exchangers. Although it is recogmzed that the test dP may be overly restrictive due to the testing configuration, it appears as if the calculations performed to determine maximum differential pressure may not have accounted for the most conservative scenarios. Based on this, an Engineering Change Request will be initiated to evaluate
- the base CDA scenmio which may require the rerunning of KYPIPE model. If it is determined that the worst case scenario was not assumed, a Deviation Report will be initiated to document the finding (Ref. Question #121.0).
Since the MOVs under question were capable of operating at the greater torque conditions, there is no safety concern with this issue.
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CORPORATE NUCLEAR SAFETY ASSESSMENT l NORTH ANNA POWER STATION SWSOPA B 04.c Preoperational test demonstration of system capability and limitations.
The pre-operational testing performed, specifically testing performed after design change implementation, was found to be acceptable.
04.d Adequacy of modification testing.
Several design changes were reviewed to determine the adequacy of post-modification testing. Generally, post-modification testing appeared to be adequate to demonstrate system and component operability. The following Design Change Packages were reviewed for adequacy of testing results:
DCP 92-123, "SW Strainer Replacement" DCP 92-266, " Replacement of SW Pump" DCP 84-31," Spray Array" DCP 91-04, "SW 4" Control Room Chiller Piping Reroute" 04.e Technical adequacy of IST procedures, trending of results and recurring failures. Completeness ofIST program.
The IST program basis was reviewed to determine whether the IST program was complete. The IST program " boundary" was generally found to be acceptable. However, several examples were found where the IST program did not include components that should have been.
This was because of their safety importance or the potential for leakage that was greater than assumed in the design analysis . They are:
IST program did not include valves located in the chemical addition subsystem to the 36" SW headers. These valves function as the boundary between the safety related SW system and the non safety related chemical addition system. A Deviation Report (N-94-985) was issued to document the finding and to identify corrective action. The ISI group will include the manual isolation valves (1-SW-1139,1-SW-1067, and 1-SW-1070) in the IST program unless an engineering analysis determines that the leakage from these valves is acceptable. The manual isolation valves will become the safety to non-safety related boundary and will continue to be maintained open to inject chemicals (Ref.
Question #002.0).
As part of the response to Question #002.0, the ISI group committed to testing the check valves in the closed direction to ,
determine whether these valves would have prevented backflow '
from the SW header into the chemical addition subsystem. The chemical addition system is configured such that each service water header is supplied with a chemical addition'line. Each Page 68 l
l I
CORPORATE NUCLEAR SAFETY ASSESSMENT -l I NORTH ANNA POWER STATION SWSOPA I j chemical addition line contains a manual valve, close to the SW header, and two check valves. These two individual chemical addition lines join into a common header. The two check valves in each individual chemical addition line were tested together since j individual testing is not possible due to configuration. The i surveillance test on these check valves determined that the check I valves supplying the "A" SW header would not have prevented SW !
backflow through the valves. Please refer to Section 04.g for_ l further discussion regarding the surveillance testing. ;
i In a similar situation, the non safety-related Polymer Addition i System lines to the Control Room Chiller SW sub-system was ;
found to be connected through normally open manual valves. The team questioned if these valves should also be in the testing i program. The response indicated these valves would be changed .
to normally closed, eliminating the need for testing (Ref. Question !
- 058.0). The corrective action was found to be acceptable.
The check valves located on the Recirculation Spray Heat i Exchanger supply header, specifically valves,1-SW-114,1-SW-116,2 SW-068, and 2-SW-070 are, according to the IST Program Basis, required to isolate a ruptured SW header. However, these valves are tested by the IST program only to verify their ability to open with no verification performed to ensure their ability to i close. If a closure test is not performed rn these valves, there is i no guarantee that they are capable of preventing backflow. !
Since the integrity of the piping is established by design, l precluding the need for the check valves to isolate as a result of piping failure, it is reasonable to not perform any closure tests on the check valves since they are not needed to prevent flow through i a ruptured header. Further, a dual unit outage would be needed !
to back flow these valves. However, this is inconsistent with !
UFSAR Section 9.2.1.3.2 that indicates these check valves serve to prevent an operable Service Water header from back-feeding an inoperable supply header. It is recommended that UFSAR :
Section 9.2.1.3.2 be revised to delete the reference to !
isolating a ruptured SW header with the header check valves (Ref. Question #013.1). -
Valve surveillance trending was found to generally be acceptable. t The current valve stroke time program considers the system design i conditions. Similarly, pump surveillance trending was found to be ,
acceptable with trending results used as required to predict future :
results. '
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CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA 04.f Calibration and testing ofinstrumentation, valve stroke time, test equipment used for TS operability. . Tolerances for instrument accuracy.
Question 93.0 was written to document the fact that during the SW pump head verification testing, the procedure allows the use of either the instrumentation located at the pump house,1-SW-CPU-101 or the bypass flow indication,1-SW-FI-110-1 to be used for the evaluation of pump performance. The flow values obtained from either of these instruments is plotted against the developed pressure difference to obtain a pump curve. The data obtained during a test indicated that the two instruments differed by as much as 700 gpm as identified in Question 93.0. Of concern was whether the tests performed over time were being compared to tests in which the came instrument was being used.
In response to the team's question, it was determined that the pump curves developed for comparison have used the results of the bypass flow instrumentation. Since the head curves are for the purpose of comparison, the difference in the results obtained from the two separate instruments is not of concern. However, sufricient CPU instrumentation data exists to develop a new reference curve for the pumps. Since assumptions and the analysis performed have to date been based on the bypass flow instrumentation, consideration for the flow differences as well as accounting for pump degradation must be taken into account prior to using the CPU data.
With respect to the diiTerences in the flow values attained from the separate instrumentation, each instrument is calibrated as required by the Calibration Program to be within the required tolerances. As indicated in the answer to Question 93.0, the differences can be as a result of differing technology. Also, the ditTerences can be as a result of physical placement of the instrumentation. Engineering appeared to be well aware of the potential issues and this was found to be acceptable.
A sample was taken to determine the acceptability of the Calibration Program. The Calibration History on several gauges were reviewed and found to be acceptable. The gauges reviewed were:
NQC-4050 NQC-4020 NQC-047 NQC-433 04.g Witness PMT, surveillance and ISTs for SW.
The review of the surveillance testing and post maintenance testing resulted in the following:
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CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA
+
The team witnessed a performance test of the chemical addition system boundary check valves that were identified by the team as safety to non safety boundary valves that were not included in the IST program (refer to Section .04.e above). The testing perfonned on the check valves determined that the check valves supplying the "A" SW header were incapable ofisolating the safety related SW header from the non-safety related chemical addition piping.
Deviation Report, N 941026, was written to document this incident. Failure of these check valves to prevent backflow may have existed for a significant period of time since these check valves have not been previously tested by the IST program.
This concern was resolved by closing the manual valves closest to the SW header eliminating the potential for future failure of the check valves to prevent backflow.
The team evaluated the post maintenance testing (PMT) performed after maintenance on valve 1-SW-22, the 1-SW-P-4 .
Auxiliary SW Pump discharge check valve. The valve was disassembled for a routine IST inspection on 3-29 94. As specified by the PMT program, a leak test of the valve after maintenance to ensure pressure boundary integrity was required. However, the PMT follow sheet specified post maintenance testing for 2-SW-22 rather than the applicable valve, resulting in 1-SW 22 being returned to service without the required testing. A Deviation Report (N-94-1018) was written to document this finding. A review of the documentation determined that a leak test and a backseat test were successfully performed on 4-22-94.
Since the affected valve was shown to operate as required in a subsequent test, any concerns were eliminated regarding the ability of tha valve to operate as required.
04.h pts for safety related heat exchanger heat transfer capacity and trending.
The team reviewed the Virginia Power responses to Generic Letter 89-13 to determine the commitments made to the NRC with respect to heat exchanger performance testing. The team also reviewed several heat exchanger performance test results. Generally, the performance tests appear to be in accordance with the commitments made and they appear to adequately assure that macrofouling as well as microfouling issues are addressed. The following are the results of the team's review:
The Recirculation Spray Heat Exchangers are maintained in " dry layup" to avoid microfouling since performance testin.g at design basis conditions cannot be performed. As part of the SW flow testing, Page 71 ,
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWEll STATION SWSOPA the flow through the heat exchangers is verified to be as required by design basis. Question 79.0 was issued to ascertain whether design conditions, specifically the time required for the heat exchangers to become water solid to ensure adequate heat transfer, were considered and verified through testing. Of interest is the time required to evacuate the heat exchanger of air and whether this time duration is less t'uan that required by accident analysis, it was found that no specific testing has confirmed the availability of the heat exchangers to fill within the time period required by design. There is evidence, however, from testing on Surry's heat exchangers, that the North Anna HXs will become water solid before the time required by design. ,
The Surry HXs are similar in design and configuration to the North Anna RSHXs. Furthermore, new V-Cone flow elements will be installed during the upcoming refueling outages on the units which will provide reliable indication ofinitial filling of the RSHXs.
The Control Room Air Conditioning Chillers heat exchangers are tested to determine their ability to perform as required by the design basia. Design basis conditions cannot be tested because normal heat loads are less than accident loads. However, an analysis is performed which extrapolates the data obtained during the test to accident conditions.
The Component Cooling Heat Exchangers are tested to determine their ability to perform as required by design requirements. Once a year one Component Cooling Heat Exchanger is tested to verify the operability of the heat exchanger. A performance test,2 PT-74A, on the 2 CC-E-1A was reviewed to determine the testing methodology and the analysis performed to ensure the operability of the heat exchanger. The team found that the test adequately demonstrated heat exchanger performance. The team also found that the conclusions reached regarding the extent of fouling to be reasonable.
The Charging Pump Lube Oil, Gear Box and Seal coolers are not performance tested. Instead the coolers are inspected and cleaned on an annual basis and charging pump parameters are monitored which provides assurance that the heat exchangers are performing as required.
04.1 Component unavailability data compared to IPE.
No specific review was performed to identify whether the component unavailability is less than that assumed in the IPE. Previous review in this area during Surry's SWSOPA did not reveal any problems.
04.j Testing of component performance in accordance with design !
basis. '
Please refer to the section 04.b above. !
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f CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA 04.k pts to detect flow blockage from biofouling in other systems.
A review of other systems having in direct contact with the service <
water system was made to determine the extent of the testing performed as well as the acceptability of the results of this testing.
The team found the testing, where performed, to adequately ensure detection of biofouling.
One area of concern was identified. The CCHX radiation monitor,1-SW-RM-107, and the SW discharge radiation monitor,1-SW RM-108, sample from both SW discharge headers simultaneously. The method of verifying flow from each header was questioned, since the flow indicator for each monitor is located downstream of where the individual sample lines combine. The response indicated that the flow paths are not verified individually. This could result in the' undetected failure to sample one of the headers, since blockage from biofouling, debris, or a check valve failure could prevent a sample from being obtained. A method to periodically verify that these lines are unblocked should be developed (Ref. Question #080.0).
04.1 Testing one air to water HX for proper heat transfer.
Examine air side for fouling.
No direct SW to air HXs exist at NAPS.
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P CORPORATE NUCLEAR SAFETY ASSESSMENT l NORTH ANNA POWER STATION SWSOPA
- 05. QUALITY ASSURANCE AND CORRECTIVE ACTION :
A. SIGNIFICANT ISSUES
- 1. Strengths
- " Repeat" DR identification SW Restoration Project Engineering support for SWS B. ASSESSMENT TOPIC DISCUSSION 05.a. SNSOC and MSRC minutes reviewed for SW item dicerennneles and unusual onerability determinations.
t Review of Station Nuclear Safety and Operating Committee (SNSOC) and Management Review Board (MRB) meeting minutes indicated ;
that SNSOC and the MRB discussed SW item discrepancies. No unusual operability determinations were noted. Due to the level of detail in Management Safety Review Committee (MSRC) meeting j minutes, they were not reviewed. The following is a list of some of the SW issues SNSOC and MRB discussed in the last six months:
- 1. Appendix 'R' action requirements relating to Auxiliary SW pumps '
and valves
- 2. Numerous issues and procedure changes related to the SW Preservation Project
- 3. Microbiologicalinfluenced corrosion
- 4. Inserviceinspection
- 5. Throttling alignment
- 6. Maintenance
- 7. Instrument calibration ;
05.b. LFRs. NPRDS. 50.72 renorts. enforcement actions. NCRs. TS onerability determinations. WOs. and adverse test results or recurrent failures. Adenunev of RCEs. l Review of LERs, NPRDS, Enforcement Actions, DRs, JCOs, and RCEs for recurrent failures did not identify programmatic problems ,
with the Corrective Action Program. One programmatic feature was i noted. When a DR is issued on a component, Station Nuclear Safety !
(SNS) has the ability to quickly identify whether a previous DR has l
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CORPORATE NUCLEAR SAFETY ASSESSMENT l l NORTH ANNA POWER STATION SWSOPA E been issued since 1990 on the same mark number and determine if. {
the deviation is similar to a previous deviation. If the deviation is the ;
same, the DR gets stamped " Repeat." When the DR is reviewed at the morning Superintendent's meeting the " Repeat" stamp results in i additional management scrutiny. This is considered a strength !
in the Corrective Action Program, i While reviewing the above documents for recurring problems, the i following specific SW items were identified as needing some !
additional attention (Note - the " Repeat" stamp process was not in !
place when the following DRs were identified): ;
SW to lake radiation monitor Radiation monitor 1-SW-RM-108 (on the SW return to the lake) has ;
had a recurring problem related to personnel inadvertently- i disconnecting the power supply. Although the station seldom uses j the lake to lake mode of the SWS, it is part of the system design. ;
Seven DRs have been written since 1990 to present, and 15 :
Corrective Maintenance (CM) work ' orders since -1984 have been 1 issued to correct the same problem. Recently, Request for !
Engineering Assistance (REA) 93 289 determined that no corrective action is necessary because activity in the area has been curtailed.
i However, personnel continue to enter the area, and there is a strong i potential that the unit will be inadvertently disconnected again without some protection for the power supply cable. It is recommended that protection for the power supply connector !
that is becoming disconnected be reconsidered (Ref. Question !
- 034.0).
Radiation monitor numos SW radiation monitoring pumps for both units have.had a history of binding problems from SW corrosion products and sediment entering 1 tight clearances in the pumps. The RSHX radiation monitor pumps i for both Units were replaced to correct this problem during the i respective 1992 and 1993 outages (DCP 90-12). However, the !
change-out of the SW radiation monitoring pumps 9A,9B and 10 was !
not performed because they were not within the scope of DCP 9012. !
To respond to DRs written after the 1993 outage that were related to :
SW radiation monitoring pumps 9A, 9B and 10, REA 93-348 was ;
written. The REA indicates that the original pump can be used with .
internals made from a different material (bronze)instead of changing !
the pumps to Gould pumps similar to the radiation monitoring !
pumps on the RSHXs.
Engineering is in the process of preparing an Item Equivalence Evaluation Review (IEER) to change out "like-for-like" components made of bronze instead of the original material. The s'upervisor of l r
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i CORPORATE NUCLEAR SAFETY ASSESSMENT !
l' NORTH ANNA POWER STATION SWSOPA l l
Procurement Engineering was contacted to determine if the IEER process was appropriate to be used instead of a DCP for a pump material change. He indicated it is usually a judgment call, and that the IEER process contained screening criteria to assist in this determination. At the time of this assessment, Procurement Engineering had not started the evaluation due to other priorities.
ReliefValves During 1989, VPAP-0804, " Safety Relief Valve Program," was implemented. In 1990 relief valves 2-SW-RV-200A, B and C, and Unit I relief valves 1-SW-RV-1000 and D failed to lift within the criteria established in the preventive maintenance (PM) work order. These valves were re-worked and placed back in service. As per VPAP-0804, the testing frequency was reduced (from 5 years to 3 years).
However, in 1991 the Unit i relief valves were tested again, and again they failed to lift within the criteria established in the PM work order. In 1993 all four of the subject Unit 2 relief valves failed to lift within the criteria established in the PM work order. It should be noted that the relief valves functioned just outside (1 or 2 psi) of the criteria established in the periodic test. Action was taken to simply re-work the valves based on the limited significance of the deviation from the as found setpoint and small likelihood these valves would be required to function during a design basis accident.
These failures have been attributed to the process fluid and the frequency of maintenance and testing. These valves are only exposed to their normal process fluid during RSHX flow testing that occurs on a refueling basis. Otherwise, they are only exposed to air. The valve's history indicates it is not assured they will lift within the design pressure range in the event they are needed. Phough there has been no code violation, action should be taken .; assure the RSHX relief valves will perform as required if they fail to meet the established criteria when they are tested (Ref.
Question #152.0).
05.c. Comoarison of SW assessment results with aunlity verification nrograms.
The second paragraph of GL 8913 references 10 CFR 50 Appendix B, which states, "The quality assurance functions consist of(a) assuring that an appropriate quality assurance program is established and effectively executed and (b) verifying, ..., that activities affecting the safety-related functions have been correctly performed." The team evaluated whether the Station Quality Verification organization has previously identified the same issues the SWSOPA Team has identified. Also evaluated were actions the Station Quality Verification Organization has taken to track, escalate and ensure implementation of the issues identified.
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CORPORATE NUCLEAR SAFETY ASSESSMENT l I NOHTH ANNA POWER STATION SWSOPA l QA has not performed an overall assessment or audit of the SWS. l Howeva, QA has previously identified SW issues similar to those identified by the team. These issues can be identified in the QA Trending and Tracking database. For example, the material condition of the SW expansion joints were identified by items I 01219, I-94-00970 and I-94-00933. Some examples ofitems related to material condition can be found associated with I 93-00432 and I 93-00493.
It should be noted the Quality Verification Organization consists of more than the Quality Assurance (QA) Department. The total r Quality Verification Program at Virginia Power consists of the efforts performed by QA, Station Nuclear Safety, Corporate Nuclear Safety and Station personnel. Each individual associated with the SWS has the written responsibility (VPAP 1501, " Deviation Reports") to report deficiencies related to the SWS. There is every indication this is occurring.
05.d. Timeliness and technical adeounev of resolution ofitems from self assessments. Onen item trackina and closure.
The timeliness and technical adequacy of corrective action related to the SW items identified by the Quality Verification Programs is evident by the general lack of new recurring problems.
The SW Preservation Project is an example of significant and massive corrective action for a long-standing problem. The technical resolution of problems includes innovations in restoring and coating SW piping and then protecting it with chemical treatment. The team considers this project a strength. ,
05.e. Interface between enaineerina/ technical sunnort and i onerations for resolution of onerational nroblems.
Based on conversation with operators on shift and the System Engineer, there is a good working relationship between Operations and Technical Support. Operational questions or problems related to the SWS are directed to the SW System Engineer. During the recent SW Preservation Project, Operations and the System Engineer worked together to maintain operability of the SWS for the operating Unit, and to address the issues related to the preservation Project. ,
The team considers this a strength.
Page 77
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA GENERIC LETTER 8913
SUMMARY
SERVICE WATER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EQUIPMENT ACTION I Implement and maintain surveillance program and controi techniques to reduce flow blockage problems due to biofouling.
Enclosure 1 to the GL 89-13 provides an acceptable program. The following addresses the individual requirements and the response to each aspect of the Enclosure.
Reauirement A - Visual inspection ofintake structure once per refueling outage for biofouling, sediment and corrosion.
Response - Letter serial number 89 572, dated January 29,1990, (initial response to GL 89-13) and subsequent update response letter serial number 89 572H, dated October 18,1991, states:
" Visual inspection of the intake structures at our nuclear stations will be performed as part of our routine inspection and maintenance program discussed in the response to Item III."
The update response further states:
"The intake structures for the auxiliary SW pumps,1-SW-P-4 and 2-SW-P-4, and two of the normal SW pumps,1 SW-P-1A and B, were inspected. The suction bowls for the auxiliary SW pumps are located about 16' above the floor of the intake bay. As a result, macrofouling by Asiatic clams is not a concern. The overall material condition of the pumps is good and there is no indication of any :
biological macrofouling. Clams were found only in the low flow regions on the '
intake bay floor. These regions were relatively small and had low clam !
population densities. Few clams were found in the normal SW pump intake bay l (restricted to low flow areas). Mud and silt had accumulated in the low flow regions of the intake structure, but most of the concrete floor was clear. No obstructions were found in the traveling screens. The inspection did not find any biological macrofouling or silt accumulation that could affect the NPSHA for the ;
normal SW pump.
Virginia Electric and Power Company instituted a program to monitor Lake Anna for zebra mussels and Asiatic clams. ... Inspections are performed during the spring and fall seasons."
Comments - A more recent inspection of the intake bay for auxiliary SW pump,1-SW-P-4, was performed by System Engineering on 3-19-93 per W.O.151333.
The inspection did not identify any macrobiological growth and the screens were found to be clear of foreign material. No Asiatic clams were found. The results of the inspection is summarized in 0-PT-75.15 completed on 8 3 93.
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] CORPORATE NUCLEAR SAFETY ASSESSMENT l
- l- NORTH ANNA POWER STATION SWSOPA E Based on the.results of the previous inspections, the frequency of intake i structure inspection is every 5 years. Corrosion of the SW traveling screens was identified in 1989 and Work Orders written to replace the screens. The condition of the screens was not documented in 0-PT-75.15. Action to correct g the condition of the screens is being taken after concerns were raised by the assessment team.
The program to monitor Lake Anna for zebra mussels and Asiatic clams has been established. The results of the Spring 1993 survey were attached to 0-PT-75.15. This survey did not identify any live clams in the SW reservoir. No zebra mussels were found in the Lake Anna samples and the number of clams collected were low compared tu previous Spring samples. No clams greater than two years of age were collected.
Reauirement R - The SWS should be continuously treated with chemicals to control macroscopic biofouling.
Response - Letter serial number 89 572, dated January 29,1990, states:
The North Anna Power Station currently employs a chemical treatment program to mitigate biofouling and corrosion problems. We have reviewed the existing program and determined that it is adequate.
Comments - Macrofouling has not been a concern in the North Anna Service Water '
system. The SW reservoir is chemically treated with biocides for MIC. !
Reauirement C - Redundant and infrequently used cooling loops should be flushed and flow tested periodically at the maximum design flow to ensure that they are not fouled or clogged. Other components in the SWS should be tested on a !
regular schedule to ensure they are not fouled or clogged. SW cooling loops should !
be filled with chlorinated or equivalently treated water before lay up. ;
Response - Letter serial number 89 572, dated January 29,1990, states:
At North Anna, periodic flushing and/or flow testing is conducted for some infrequently used service water loops. For those component loops that are currently flushed or tested, appropriate procedures will be reviewed and revised as necessary to ensure that minimum design flows are achievable. Applicable performance tests will also be revised to require inspections of associated service water lines if biological fouling is indicated. In other cases, flow testing of infrequently used cooling loops is not practical. This is particularly true in those cases where service water acts as a backup system and introduction of raw water to the primary cooling system or water source would be undesirable. As an alternative, inspection procedures will be developed to provide assurances that infrequently used cooling loops are not significantly fouled. North Anna does not have a history of biofouling causing flow blockage in the service water system.
The update response of October 18,1991 further states:
- Page 79
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA Preventive Maintenance Procedures (PMs) for the charging pumps were upgraded to include annualinspections for the seal coolers. PMs were created to inspect and clean, if necessary, the Component Cooling Heat Exchangers (CCHXs) on an annual basis. No change was needed regarding the existing annual inspection of the control room chiller condensers... -
The Inservice Testing Program at North Anna requires full flow testing of SW lines which have check valves in the program. Such check valves are found in .
the SW lines associated with the RSHXs, SW pumps, charging pumps, and control room chillers. These tests provide adequate indication of the absence of '
blockage in the lines.
Comments - Annual inspections and cleaning of the control room chillers and the charging pump coolers are being performed as indicated by review of completed work orders.
NDE measurements are taken on one of the two supply lines and one of the two return lines to the fuel pit coolers. These lines provide SW backup to the fuel pit coolers which are normally :ooled by Component Cooling water. These lines are not routinely flushed. The NDS results indicate that the corrosion rate is low and the wall thickness of these sampled lines is well above the minimum wall thickness.
Service Water lines to the Auxiliary Feedwater (AFW) Pump suction are maintained dry (the lines tap off the supply lines to the RSHXs downstream of MOV-SW-101s which are maintained dry during normal operation) and flushing of the AFW supply lines is therefore not required.
Reauirement D - Samples o." water and substrate should be collected annually to determine if Asiatic clams have populated the water source.
Response - See response under Requirement A above.
Comments -See comment under Requirement A above.
ACTION II Conduct a test program to verify the heat transfer capability of all safety related heat exchangers cooled by service water.
The total test program should consist of an initial test program and a periodic retest program. In lieu of a test program, frequent regular maintenance of the heat exchanger is an acceptable alternative.
Response - The update response of October 18,1991 states:
Periodic Tests are performed on a routine non-outage basis to verify that control room chillers and CCHXs meet design requirements for heat removal. These tests demonstrate that preventive maintenance provides for adequate performance. The surveillance frequency will be establishe'd following Page so
CORPORATE NUCLEAR SAFETY ASSESSMENT l I NORTH ANNA POWER STATION SWSOPA E I performance of several tests. Test have been performed on the control room chillers and CCHXs Preliminary test results indicate design performance in all i tested heat exchangers. ;
I Heat transfer testing of the RSHXs is not performed. The system physical I configuration makes heat transfer testing impractical without extensive ;
modifications. The RSHXs are maintained in dry layup during normal operation in order to preclude the possibility of microfouling. Periodic testing is performed which verifies adequate SW flow is available to the RSHXs to support post DBA i containment depressurization and long-term containment heat removal. ,
i Comments - In addition to the above actions, the SW side of the control room chillers and the charging pump coolers (lube oil cooler, gear box cooler, and seal coolers) are ,
inspected and cleaned,if required, on an annual basis. The type and quantity of debris in the coolers has not been well documented in the past but station personnel l indicated that flow through the coolers is not restricted. The coolers are normally cleaned each year, even if minimal debris is found. Component performance is also i monitored to ensure that the equipment is operating satisfactorily when in service. l Technical Report No. ME-0025, Rev.1, "NRC Generic Letter 89-13 Activities,"
indicates that design margin allows for degradation of the heat transfer capability :
of these heat exchangers. !
1 The spent fuel pit coolers are normally cooled with component cooling water which !
is chemically treated, therefore the heat exchangers are maintained in a clean !
condition. SW is available as a backup if needed. l i
The testing methodology andjustification are further discussed in Technical Report No. ME-0025 for each of the heat exchangers covered by GL 89-13.
ACTION III Maintenance and Inspection Program for SW piping and {
components to ensure no corrosion, erosion, coating i failures, silting, and biofouling that degrades the performance is present.
Response - Letter serial number 89 572, dated January 29,1990, (initial response to GL 8913) stated: j Various activities exist at both North Anna and Surry Power Station that constitr.te elements of an inspection and maintenance program. As noted in our :
respot.se to item II, various heat exchangers are currently subject to inspection :
and periodic cleaning. We also will inspect the service water intake structures and those additional pipe sections which provide assurance that infrequently used cooling loops are not significantly fouled. Accordingly, we will review our current activities and organize them into an integrated inspection and maintenance program. Additional areas ofinspection will be evaluated on a case by case basis. Our inspection and maintenance program will be developed prior to restart following the next refueling outage for each respective unit. '
5 Page 81
~ CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA Concerning inspection of the Spent Fuel Pit Coolers, the update response of April 30,1991 further states:
Inspection of(1) CCHX for Biofouling will ensure absence of macrofouling in this portion of the SW System. SW is backup cooling source for these coolers. SFP HX lines are not inspected. CCHX provides proxy for inspection.
The update response of October 18,1991 further states:
Component Cooling Heat Exchanger 1-CC-E-1A was inspected before and after cleaning. The as found condition of the CCHX was good with no trash or debris present. About ten clams were found inside the CCHX, and approximately ten tubes could not be hydrolased due to blochage. Performance testing has indicated that this condition does not affect the ability of the CCHXs to remove design basis heat loads. Some mud was found in the lower head when the CCHX was opened. Minor corrosion was also found in the carbon steel associated with the CCHX heads. Hydrolasing was effectively used to clean the tubes and tubesheets of deposits. The overall as-found conditions of the Unit 1 CCHXs are considered to be identical to those of the Unit 2 CCHXs. Ongoing performance testing will be used to determine the required cleaning frequency for the CCHXs.
The current frequency is conservative based on observed heat exchanger performance.
Comments - Virginia Power's updated response of April 30,1991 indicates that inspection for the lines that supply the spent fuel coolers will be based on the CCHX inspection. However, the criteria in 0 PT-75.15, " Generic Letter 89-13 Service Water System Testing Requirements," to initiate this inspection of the SW lines that supply the spent fuel coolers is in step 6.1.6 which states, " Review GL 89-13 and Virginia Power responses to GL 89-13 to ensure that all commitments are met."
This PT has been performed and to date Virginia Power has not inspected the SW lines that supply the spent fuel coolers. The Station has stated that the reason the lines have not been inspected is because the CCHX have not been found fouled enough to initiate the inspection.
Virginia Power's updated response to the NRC indicates that CCHX 1 CC E-1A had some degree of fouling. The response states that the blockage in ten of the tubes could not be cleared with a hydrolaser. This indicates that there may have been many more tubes that had blockage and were cleared with the hydrolaser. Based on the criteria 0-PT-75.15, it appears there was sufficient fouling in the CCHX to warrant an inspection of the SW lines that supply the spent fuel coolers. Note there is also no criteria to inslect i the SW lines that supply the spent fuel coolers in procedure 0 MCM 080101, " Cleaning Component Cooling Heat Exchangers." To ensure that the specific commitment to inspect the SW lines that supply the spent fuel coolers is not inadvertently missed after future inspections, it is recommended that 0 PT 75.15 he revised to incorporate the commitments to inspect the SW lines to the spent fuel coolers made in the latest NRC ;
response to GL 8913 (Ref. Question #126.0). ,
l l
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I CORPORATE NUCLEAR SAFETY ASSESSMENT l NORTH ANNA POWER STATION SWSOPA __
l In addition, VPAP 0811, " Service Water System Inspection and Maintenance Program" requires an inspection of the SWS pipe for protective coating failures.
Some SW piping has been recently replaced with protective coating, and as indicated in a clarification of the Preservation Project (Fax date 7/26/94), initial inspection of the protective coating after it was in service for six months indicated that the two 100% coating system is working effectively. The clarification statement also indicated that a protective coating failure inspection program will be evaluated at the end of the project. Therefore, it does not appear the Station is implementing the requirement to perform a routine protective coating failure inspection. GL 8913 states such a program should be established before plant start-up following the first refueling outage beginning 9 months after the date of the generic letter (7/18/89).
Protective coating is also applied to other SW components. For example the main SW pump columns and screen wash pumps have a protective coating. The screen wash pumps have noted as having degraded coating.
It is recommended a program be established to inspect the protective !
coating of SW piping and components (Ref. Question #086.0). -
NDE is performed on selected, non coated piping in the SW system. The piping which is monitored is identified in 0-PT-75.14, Service Water Wall Thickness Monitoring, which is performed on an annual basis. Corporate Project Engineering then evaluates the data taken to determine corrosion rates and recommends piping repairs / replacement when deemed necessary.
ACTION IV Confirm the SWS will perform its intended function in accordance with the licensing basis for the plant. This confirmation should include the ability to perform required safety functions in the event of a single active failure. To ;
ensure that the as built system is in accordance with the l appropriate licensing basis documentation, this '
confirmation should include recent (within the past 2 years) i system walkdown inspections. Results of the single failure analysis and plant walkdown should be maintained in Station records. ,
i Response Letter serial number 89-572, dated January 29,1990, (initial response to GL 8913) stated:
Virginia Electric and Power Company is currently undertaking an extensive Configuration Management Project. Included in this effort is establishing the design basis for the SWS. The resulting Design Basis Documents and associated references will be reviewed to confirm that the SWS will perform its intended >
safety function. The above review will also ensure that safety functions of the ,
SWS are not vulnerable to a single failure of an active component.
The update response of October 18,1991 further states:
Page 83
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA A specific review of the SW DBD was conducted, and this review concluded that any single failure would not prevent the system from performing its design function. A physical walkdown of the SW system was performed as part of the design basis verification program to ensure that the construction of the SW system conforms to design documents.
Comments - Virginia Power's initial response stated that the above required walkdown was performed as part of the Configuration Management Project.
Although no significant discrepancies were found during this assessment that would prevent the system from performing its safety function, the Operations section of this report discusses labeling and drawing discrepancies.
The Station did perform a single failure review, however, the documentation of that review is marginal and during the walkdown some additional single failures were identified. These issues are discussed in the Engineering Section of this report.
ACTION V Confirm that maintenance practices, operating and emergency procedures, and training that involves the SWS are adequate to ensure that the safety related equipment cooled by the SWS will function as intended and that operators of this equipment will perform effectively. This confirmation should include recent (within the past 2 years) reviews of practices, procedures and training modules. The intent of this action is to reduce human errors in the '
operation and repair of the SWS. The results of this review should be documented in station records.
Response Letter serial number 89-572, dated January 29,1990, (initial response to GL 89-13) stated:
As the design basis documentation is completed, we intend to review our maintenance, operating, and emergency procedures, and training, to ensure that they are consistent with the design basis. ...
In addition to the programs described above, both stations are involved in a procedures upgrade program to review and upgrade the station procedures. This program will provide further assurance of the adequacy of service water procedures. Completion of the procedure upgrade program is planned for 1995.
The update response of October 18,1991 further states:
Virginia Electric and Power Company has an ongoing procedure upgrade program that requires periodic reviews of existing procedures. This is adequate to ensure that SW operation, maintenance, and test procedures are revised in a timely fashion to address any new issues. Many procedure changes have been made and SW system training was given to operating license personnelin the Licensed Operator Requalification Training Program during 1991. Similar training was conducted for engineering personnel. The procedure changes and Page 84
i l
CORPORATE NUCLEAR SAFETY ASSESSMENT I NORTH ANNA POWER STATION SWSOPA
)
l training included lessons learned while evaluating the SW DBD.. However, additional SW DBD reviews are required. As a result of 1991 North Anna 1 outage activities, the rescheduling of the North Anna Unit 1 and 21992 refueling outages, which are scheduled to occur from February 1992 through June 1992, and the issuance date of the SW DBD relative to these activities, it is necessary to revise our commitment for completing procedure and training reviews associated with the SW DBD. Therefore, SW procedures and training program reviews, associated with the SW DBD, are expected to be performed, and necessary revisions completed, within 6 months of completion of the 1992 refueling outages.
Comments - Some of the DBDs have been issued and an integration review has been performed. The Engineering, Operations and Maintenance sections of this report provide additional details related to Action V of GL 89-13.
Operations does not consider the main SW pump screens as required for operability.
even though they are part of the design basis. Emergency Operating Procedures do not monitor SWS parameters and Alarm Response procedure lack specific details.
In addition, maintenance personnel training and procedural criteria related to quantifying the amount of fouling in a heat exchanger or section of pipe is insufficient. In addition, when personnel have performed these inspections they have not written deviation reports. The Operation and Maintenance issues are addressed in greater detail in those respective sections of this report.
4 Page 85
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA ATTACHMENT 1 NORTH ANNA SWSOPA ASSESSMENT TEAM MEMBERS Team Member Location / Dent. Assignment Mike Surface INS / CNS Team Leader Francis Terminella INS /CNS Assistant Team Leader, Maintenance John Lewis INS /CNS Maintenance, Stuart Klein (Lead Mechanical Consultant Design Design Engineer)
Steve Jarema (Mechanical Consultant Design Design Engineer)
Joel Kelly SPS / System Eng. Design Jim Roth INS / CNS Design l
Tom Kendzia INS / CNS Operations Greg Prescott (SRO) SPS / OPS Operations Neil Turner SPS/ OPS Operations Manuel Bandeira (Operations Consultant Surveillance and Testing Engineer) .
Rick Bailey SPS / IST Eng. Surveillance and Testing l David Wootten INS /CNS Corrective Action Kenny Pier SPS/QA Corrective Action l
l ATTACHMENT 1 Page 86
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA ATTACHMENT 2 RECOMMENDATIONS FOR OPEN CNS SWSOPA ITEMS i i
Question #004.4 Operability requirements of SW traveling screens Reference - Section 02.h Recommendation - The design function of the screens should be reassessed, either ;
downgrading them to non-safety, removing them, or requiring them to be operable.
Question #010.1 SW pump periodic inspections Reference - Section 03.e, Topic: " Auxiliary and Main Service Water Pumps" Recommendation - The team recommends that the PMs for disassembly, inspection, and refurbishment for each of the main and auxiliary SW pumps be planned and performed in accordance with the requirements of the RCM Program.
Question #013.1 IST of SW check valves to RS HXs Reference - Section 04.e Recommendation - It is recommended that UFSAR Section 9.2.1.3.2 be revised to delete the reference to isolating a ruptured SW header with the
Question #017.1 Calibration / visibility of SW local level indicator Reference - Section 02.e Recommendation - It is recommended that the ruled marker mounted on the side of the SW pump house be verified as being accurate and a plan be developed to ensure it can be read below normal water level.
Question #034.0 01 SW.RM DR trend Reference - Section 05.b, Topic: "SW to Lake Radiction Monitor" Recommendation - It is recommended that protection for the power supply connector that is becoming disconnected be reconsidered.
Question #048.1 CR chiller inspection Reference - Section 03.d, Topic: " Heat Exchangers" and Section 03.h Recommendation - 1) Procedures used to perform inspections of HXs for the purpose of GL 89-13 compliance should be revised to require appropriate documentation of the as found conditions. 2) It is recommended that special training be provided for personnel who perform inspections of components covered by GL 89-13. Topics should include GL 89-13 requirements and commitments, recognition of asiatic clams / zebra mussels, and expectations for documenting as found conditions.
Question #049.0 SWS file in station records Reference - Section 03.i Recommendallen - It is recommended that the station comply with VPAP-0811.
Question #069.0 Qualification of MOVs & RMs to RS HX Reference - Section 01.a, Topic: " Expansion Joint Qualification" Recommendation - It is recommended that the SW expansion joints in the Quench Spray pump house basement be evaluated for radiation effects.
ATTACHMENT 2 Page 87
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA ATTACHMENT 2 RECOMMENDATIONS FOR OPEN CNS SWSOPA ITEMS (cont.)
1 Question #071.0 Accessibility & isolation of MOVs to RS HXs Reference - Section 01.a, Topic: " Service Water Radiation Monitor Sensitivity" Recommendation - It is recommended that the apparent discrepancy between the EQ zone source terms and the source term used for the shielding calculation for the RSHX radiation monitors be resolved.
Question #072.0 SW flow model questions Reference - Section 01.e, Topic: " Flow Balancing" & Topic: " Pump Runout" Recommendation - 1) The issues identified in Section 01.e under topic " Flow Balance" concerning cumulative effects of losses through non safety related, unisolated piping should be evaluated for potential impact on SWS flows 2) The basis for the SW pump composite curve used and the methodology used to develop the curve should be documented. Potential degradation should also be accounted i for. 3) The effects of maximum SWS pump flow and maximum reservoir temperature on NPSH in the " Strong" Pump /" Weak" Pump scenario should also be evaluated for impact.
Question #073.0 SW chem addition line break inventory loss Reference - Section 01.h, Topic: "SW Chemical Addition System Modification" Recommendation - Procedures should be revised in order to perform the daily l monitoring of the chemical addition system non-seismic piping added by DC 85-48 3 as required by the safety analysis.
Question #076.0 PT acceptance of SW screens Reference - Section 03.f, Topic: " Service Water System Traveling Water Screens" Recommendation - PT-75.4 should be revised to include clear acceptance criteria for the inspection of the screens for erosion and corrosion.
Question #080.0 CC HX RM flowpath and alarm Reference - Section 04.k Recommendation - A method to periodically verify that the individual header sample lines to 1-SW-RM-107 and 1-SW-RM-108 are unblocked should be developed.
Question #081.0 DBD flow criteria vs. flow balance Reference - Section 01.c Recommendation - If credit is taken for isolating SW to the RSHXs during the course of the DBA, procedures should be revised to include directions for these actions.
Question #086.0 GL 8913 action for protective coating Reference - GL 89-13 Summary, Action III Recommendation - It is recommended a program be established to inspect the protective coating of SW piping and components.
ATTACHMENT 2 Page 88
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA ATTACHMENT 2 RECOMMENDATIONS FOR OPEN CNS SWSOPA ITEMS (cont.)
Question #090.0 Cathodic protectio.1 for SWS Reference - Section 03.i, Topic: " Service Water Pump / Coating Inspection" Recommendation - It is recommended that the operation of the cathodic protection system be reviewed and the effect of any new protective coating on affected SW piping on cathodic protection design be considered.
Question #091.0 SW to SFP HXs Reference - Section 01.c Recommendation AP 27 should be revised to require performance of 0-OP-49.6 when SW is placed in service or reinoved from service for the SFP.
Question #095.0 Single failure vulnerability from AFW valves Reference - Section 02.d Recommendation - It is recommended that keeping the SW to AFW crosstie valves open be evaluated considering the impact on RSHX isolation.
1 Question #097.0 TEI recommendations Reference - Section 01.f Recommendation - The TEI recommendations for flushing sump pumps and floor drains should be implemented. 1 Question #099.0 Labels on MCR indicators Reference - Section 02.a i
Recommendation - The existence of control room " stick-on" markings that are outside the operator aid program, and the specific differences for the SW pump i discharge pressure compared to the operating procedures should be evaluated. !
Question #100.0 Single failure review issue Reference - Section 01.d Recommendation - It is recommended that components in the SW flowpath when it is the designed back up cooling source be included in the single failure review.
Question #102.0 Flooding in SW valve house Reference - Section 01.f Recommendation - The SW Valve House should be evaluated to the same criteria used by TEI to ensure that flooding does not adversely impact any safety related equipment in the area.
Question #113.0 OP 21.1 comments Reference - Section 02.b, Topic: " Specific Procedure Comments" Recommendation - The need for additior al controls when SW is aligned to the i
Containment Air Recirculation Coolers and the SW headers should be evaluated.
l ATTACllMENT 2 Page 89
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CORPORATE NUCLEAR SAFETY ASSESSMENT F NORTH ANNA POWER STATION SWSOPA j ATTACHMENT 2 ;
RECOMMENDATIONS FOR OPEN CNS SWSOPA ITEMS (cont.) !
Question #118.0 AP 5 comments Reference - Section 02.b, Topic: " Specific Procedure Comments" Recommendation - It is recommended that precautions about the potential for a sudden increase in radiation in the Quench Spray basement be included in 1/2-AP-5.
Question #119.0 Corrosion in RS HXs Reference - Section 03.d, Topic: " Recirculation Spray Heat Exchangers" Recommendation - The RSHXs should be visually inspected periodically to verify their cleanliness.
Question #126.0 Inspection of SFP cooler lines Reference - GL 8913 Summary, Action III Recommendation - It is recommended that 0-PT-75.15 he revised to incorporate the commitments to inspect the SW lines to the spent fuel coolers made in the latest NRC response to GL 89-13.
Question #143.0 Inspection of SFP cooler lines Reference - Section 02.b, Topic," Procedure Review Results" Recommendation - It is recommended that ARP IK-D4 be revised to include precautions about entering a high radiation zone.
Question #145.0 Inventory loss Reference - Section 01.e, Topic: " Flow Balance" Recommendation - Procedures should be revised in order to perform the daily monitoring of the chemical addition system non-seismic piping added by DC 85-48 3 as required by the safety analyais.
Question #151.0 Accounting for flow to non normalloads Reference - Section 01.c Recommendation - See Recommendation for Question #081.0 above.
Post Assessment Recommendations Question #152.0 RSHX Relief Valves Reference - Section 03.f, Topic: "RSHXs Tube Side Relief Valves" & Section 05.b.,
Topic: " Relief Valves" Recommendation - 1) It is recommended that piping to each RSHX relief valve be inspected and/ or cleaned. The need to inspect and/or flush these lines after they are exposed to SW flow prior to reinstallation of relief valves should be evaluated.
- 2) Action should be taken to assure the RSHX relief valves will perform as required if they fail to meet the established criteria when they are tested.
ATTACHMENT 2 Page 90
CORPORATE NUCLEAR SAFETY ASSESSMENT I NORTH ANNA POWER STATION SWSOPA ATTACHMENT 2 RECOMMENDATIONS FOR OPEN CNS SWSOPA ITEMS (cont.)
Question #153.0 Outstanding DCRs Reference - Section 02.a Recommendation - Outstanding DCRs should be processed in a timely manner.
Question #154.0 Reservoir Performance Analysis Conservatism Reference - Section 01.a, Topic: " Thermal Performance of Service Water Reservoir and Sprays" Recommendation - The conservatism in the Reservoir Performance Analysis should be confirmed considering the items discussed in Section 01.a, Topic: " Thermal Performance of Service Water Reservoir and Sprays."
ATTACHMENT 2 Page 91
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTII ANNA POWER STATION SWSOPA ATTACHMENT 3 NAPS SWSOPA QUESTION LOG R A TO: CNS# Orig. DESCRil'I' ION N N -
001.0 TK SOV drain valves in lesson plans v 4 -
002.0 MB Chemical addition check valves in IST program V -
003.1 003.0 MS Low TRC values in SWS N N -
003.1 MS Corrective Action for SW TRC N - 004.1 004.0 NT Calibration of SW PI 115B N -
004.2 004.1 NT 1-SW-PI-115B calibration N -
004.4 004.2 NT 1-SW-PI-115B calibration 4 N -
004.3 NT Log discrepancies N -
RI'I' 004.4 TK Operability requirements of SW traveling screens N N -
005.0 NT Open door at SW pump house V V -
006.0 SK Heater in SW pump house 4 -
007.1 007.0 SK SW DBD open items N N -
007.1 SK DBD open items and issues V V -
008.0 DW Old SW return lines v N -
009.0 JL P4 piping missile shields N -
010.1 010.0 JL Maintenance on SW pumps RPT 010.1 JL SW pump periodic inspections N N -
011.0 JL Corrosion coupon procedure v N -
012.0 JL Degradation of SW traveling screens 4 -
013.1 013.0 MB RS HX check valve testing N -
RPT 013.1 MB IST of SW check valves to RS HXs N N -
014.0 JK Fail position of CC trip valves to coolers N N -
015.0 SJ TDS effect on spray performance '
N N -
016.0 MS SWS concentration cycles N -
017.1 017.0 GP Guidance for operation of SW air compressor N -
Rl'I' 017.1 GP Calibration / visibility of SW local level indicator N -
018.1 018.0 NT Low temperature alarrn for SW house ;
N N -
018.1 TK Low temperature alarm for SW pump house N N -
019.0 NT Breaker load list !
N N -
020.0 TK Scaffolding on roof of SW pump house N N -
021.0 TK Housekeeping in SW pump house basement I N N -
022.0 TK Label on breaker in SW pump house N N -
023.0 NT Drawing discrepancies 11715 FM-72H, Sht.1 N N -
024.0 NT Labeling discrepancies 11715-FM 72H, Sht.1 N -
025.1 025.0 NT Drawing discrepancies 11715-FM-78A, Sht.1 N N -
025.1 NT Temporary vent & drain lines to SW piping v N -
026.0 NT Labeling discrepancies 11715-FM-78A, Sht.1 N N -
027.0 NT Drawing discrepancies 11715 FM-78A, Sht. 2 N N -
028.0 NT Labeling discrepancies 11715-FM-78A, Sht. 2 N N -
029.0 NT Drawing discrepancies 11715-FM 78A Sht. 3 ATTACilMENT 3 Page 92 l
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CORPORATE NUCLEAR SAFETY ASSESSMENT I I NOllTH ANNA POWEllSTATION SWSOPA ATTACHMENT 3 CNS NAPS SWSOPA QUESTION LOG (cont.)
l R l A l TO: l CNS #1 Orig.llDESCRIP'I'lON l N N -
030.0 NT Labeling discrepancies 11715-FM-78A, Sht. 3 4 - 031.1 031.0 JR Log SW pump discharge pressure N N -
031.1 JR Procedure update timeliness N - 064.1 032.0 MB SW pump check valve testing 4 4 - 033.0 SK Missile protection of spray arrays i -
RPT 034.0 DW 01 SW-RM DR trend N -
100.0 035.0 JK Single failure of relay N N -
036.0 JK RS HX crosstic potential N N -
037.0 JK Single CC valve misposition N -
038.1 038.0 TK SW pump casing vents N N -
038.1 TK Design evaluation of Q 38 N 4 -
039.0 GP Labeling discrepancy 11715 FB-040D, Sht. 2 N N -
040.0 GP Labeling discrepancy 11715-FB-040D, Sht.1 N N -
041.0 GP SW line-up to AFW pumps in AP 22.5 N N -
042.0 GP Discrepancy on 11715-FB-040D, Sht. 2 N N -
043.0 GP Discrepancy on 11715-FB-040D, Sht.1 N N -
044.0 GP Discrepancy on 11715-f B-040D, Sht.1&2 N -
154.0 045.0 SJ Spray pond calculation questions on ME 062 N -
154.0 045.1 SJ SW reservoir minimum spray efficiency N N -
046.0 SK SW pump discharge throttling N 4 -
047.0 SK SW pump runout basis N -
048.1 048.0 FT Inspection of CR chiller coolers N - RPT 048.1 FT CR chiller inspection N -
RPT 049.0 FT SWS file in station records N N -
050.0 TK Vent line over MCCs N N -
051.0 TK Drain line over chillers 4 -
052.1 052.0 GP Monitoring of tell tale valve 1-FW-228 N N -
052.1 GP Checking for tell-tale leakage N N -
053.0 JK Polymer addition system open valves N N -
054.0 JL Source of SW pump clearance specs.
N -
086.0 055.0 JL Pipe coating inspection v N -
056.0 JL Inspection for min. wall thickness N N -
057.0 GP Throttling position of SW pump valves N N -
057.1 GP SW pump throttle valve positions N N -
058.0 MB Testing Polvrner Addition System isolation valves N N -
059.0 JL Compatibility of SW chemicals w/ expansion joints N N -
060.0 JL Basis for 10 year expansion joint replacement N - 154.0 061.0 SJ Basis for SW heat loads N - 062.1 062.0 TK Testing of MOVs which discharge to CW tunnel N N -
062.1 TK Leak testing SW loop isolation valves ATTACHMENT 3 Page 93
l CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA ATTACHMENT 3 CNS NAPS SWSOPA QUESTION LOG (cont.)
l R l A l TO: l CNS 4 Orig.llDESCRil'I' ION l N 9 -
062.2 TK Leak testing MOVs to discharge tunnel 4 - 063.1 063.0 DW Power supplies to SW RMs '
N N -
063.1 DW Power supplies to SW RMs 1 4 - 064.1 064.0 TK Throttling to maintain 58 psig in SWS N - 072.0 064.1 SK Throttling SW valves 4 -
065.1 065.0 JK UFSAR discrepancies N N - 065.1 JK UFSAR discrepancies ,
N N -
066.0 JK SR/NSR status of CC HX rad monitor and piping 4 -
121.0 067.0 MB MOV program acceptance criteria N N -
068.0 MB 25 psid limit on CC HX N -
Rl'I' 069.0 TK Qualification of MOVs & RMs to RS HX N - 095.0 070.0 TK Accessibility & isolation of MOVs to RS HXs N -
RPT 071.0 TK RM sensitivity for RS HXs l N -
RI'I' 072.0 SK SW flow model questions l N -
RI'I' 073.0 SJ SW chem addition hon breakinventoryloss N N -
074.0 DW Number of spray nou.les required N N -
075.0 JL SW pump house settlement & pump alignment N N -
076.0 JL I'I' acceptance of SW screens N N -
077.0 JL WO acceptance of SW screens N N -
078.0 RB Flow balance calibration data N N -
079.0 MB Filling RS HXs from dry condition N -
RI'I' 080.0 JK CC HX RM flowpath and alarm N -
RI'I' 081.0 JR DBD flow criteria vs. flow balance N N -
082.0 NT 11715-FM-78C Sht. I drawing discrepancy N N -
083.0 NT .I :beling of CH gear box coolers v N -
084.0 NT E 1715-FM-78B Sht. 3 labeling discrepancies N N -
085.0 NT 11715-FM-78G Sht. 2 drawing discrepancy N -
Rl'I' 086.0 DW GL 8913 action for protective coating N N -
087.0 MB Lower pump speed at minimum EDG frequency N N -
088.0 NT Abandoned label on drain valves N N -
089.0 NT 11715 FM-78B Sht. 3 drain valve discrepancies N -
RI'l' 090.0 FT Cathodic protection for SWS N -
RPT 091.0 TK SW to SFP HXs N N -
092.0 NT SW valve pit level alarm N N -
093.0 MB SW pump head curve verification N - 094.1 094.0 GP Flooding to Aux building from SW valve pit N2 N -
094.1 GP ARP for turbine building valve pit N2 N -
RI'I' 095.0 TK Single failure vulnerability from AFW valves s N -
096.0 MB Control of CC HX dP ATTACHMENT 3 Page 94
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA ATTACHMENT 3 CNS NAPS SWSOPA QUESTION LOG (cont.)
l R I A l TO: l CNS 4 Orig.IlDESCRil'I'lON l V -
RPT 097.0 JR TEl recommendations N N -
098.0 JR Backflow preventers N -
RPT 099.0 NT Labels on MCR indicators V -
RPT 100.0 JK Single failure review issue N2 N -
101.0 TK 0-OP-49.6 comments N -
RPT 102.0 JR Flooding in SW valve house N N -
103.0 MB Flow curve basis N N -
104.0 NT RS HX drain valves and containment integrity N N - 105.0 FT Torque requirements N N -
106.0 JK Basis of SW reservoir high level setpoint V N -
107.0 MB Testing ASW pumps at design N N -
108.0 JL Maintenance of CC/SW interface valves V N -
109.0 JL SW pump relay maintenance frequency -
N2 V -
110.0 TK Checking SW parameters in an accident N N -
111.0 FT PM for IA compressor HXs N N -
112.0 MB Flow balance dP across RS HXs N -
HPT 113.0 TK OP-21.1 comments N 4 -
114.0 JL SW pump motor maintenance N N -
115.0 JK Basis for ASW pump alarm setpoint N N -
116.0 TK 0-OP-49.1 comments N2 N -
117.0 GP AP 5.1 comments N2 - RI'I' 118.0 GP AP 5 comments N -
RI'I' 119.0 JL Corrosion in RS HXs N N -
120.0 TK 1-OP-48.1 comments N N -
121.0 MB dP and torque requirements N N -
121.1 MB MOV design dP N N -
122.0 TK OP-49.4 comments N N -
123.0 NT Numbering for flow instrumentation N N -
124.0 JK Setpoint for SW return header flow N N -
.25.0 JK Basis for ASW high pressure setpoint N -
RlYI' 126.0 DW Inspection of SFP cooler lines N N -
127.0 JR Draining header for freeze protection N2 N -
128.0 GP AR 1E-F5 and flooding N N -
129.0 TK OP-49.3 comments N N -
130.0 TK OP-49.2 comments l N N -
131.0 TK 0 AP-12 comments v N -
132.0 SK SW temperatures N2 N -
133.0 NT ARP 1K-G5 for loss of air at SWPH l N2 N -
134.0 NT ARP IJ B4 N2 N -
135.0 NT ARP IJ E5 ATTACHMENT 3 Page 95
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA ATTACHMENT 3 CNS NAPS SWSOPA QUESTION LOG (cont.)
l R l A l TO: l CNS 4 Orig.ll DESCRIPTION l N2 N -
136.0 NT ARP IJ D4 & IJ-E3 N2 N -
137.0 NT ARP IJ-D3 V V -
138.0 NT 1-AR-1G-Al locating CC leakage N2 N -
139.0 NT ARP IJ B3. SW leaks y N -
140.0 NT ARP IK-D2 42 N -
141.0 NT ARP1EC7 N2 N -
RI'I' 143.0 NT ARP 1K D4 N2 V -
144.0 NT ARP IJ H6 N -
RI'I' 145.0 JR Inventory loss N N -
146.0 JR Revisions needed due to DC 92-123 N N -
147.0 JR Revisions needed due to DC 85-48 v2 N -
148.0 NT ARP IJ-D5 N N -
149.0 NT ARP IJ.G7 V -
154.0 150.0 SK Core uprate reflecting in heat loads v -
RI'l' 151.0 JR Accounting for flow to non-normal loads Lecend R = Response received V2 = Second response received A = Response accepted TO: = Indicates some other question or the report addresses the question RPT = Issue addressed in the report, including a recommendation for resolution i l
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ATTACllMENT 3 Page 96 l
l CORPORATE NUCLEAR SAFETY ASSESSMENT l _ NORTH ANNA POWER STATION SWSOPA ATTACHMENT 4 ASSESSMENT PLAN OBJECTIVE:
To conduct assessments of the North Anna Service Water System to verify that the system design, operation and performance meets design basis and regulatory requirements.
The assessment will be conducted accordance with NRC Temporary Instruction (TI) for Service Water System Operational Performance Inspections (SWSOPI), TI 2515/118, Revision 1 and Inspection Procedure (IP 40501), " Licensee Self-Assessments Related To Area Of Emphasis Inspections,"in support of Level 1 item 845 01.
SCOPE:
The Service Water System Operational Performance Assessment (SWSOPA) to be conducted at North Anna will meet the requirements of the NRC's Temporary Instruction for Service Water System Operational Performance Inspections, TI 2515/118, Revision 1. This requires verification of:
- 1. Thermal and hydraulic performance in the operation of all Service Water System equipment and configuration control in accordance with the engineering design bases (TI-01.02).
- 2. Operational controls, maintenance, surveillances, testing, Quality Assurance, I corrective actions, and personnel training to assure the Service Water Systems are capable of performing its safety-related function (TI-01.03).
j
- 3. Actions taken in response to Generic Letter 89-13 (TI-01.01).
SCHEDULE:
- 1. One week ofin-oflice preparation and team training, July 5 - July 8.
- 2. Three weeks at North Anna, July 11 - July 29.
- 3. A draft report to be available for comment August 26.
- 4. Final report to be issued September 23. l
REFERENCES:
ATTACllMENT 4 Page 97
CORPORATE NUCLEAR SAFETY ASSESSMENT I NORTH ANNA POWER STATION SWSOPA ATI'ACHMENT 4 ASSESSMENT PLAN (cont.)
- 2. Generic Letter 89-13, " Service Water System Problems Affecting Safety-Related Equip."
- 3. Generic Letter 89-13, Supplement 1-4/4/90
- 4. GL 89-04 !
- 5. IN 90-39,"Recent Problems With Service Water Systems."
- 6. IN 90-26, " Inadequate Flow of Essential Service Water to Room Coolers & Heat Exchangers for Engineered Safety Feature Systems," 4/24/90
- 7. IN 92-49, "Recent Loss or Severe Degradation of Service Water Systems," 7/2/92
- 8. NUREG/CR-5865, " Generic Service Water System Risk-Based Inspection Guide," 5/92
- 9. Temporary Instruction 2515/118, " Service Water System Operational Inspection (SWOPI)," 12/29/93
- 10. Temporary Instruction 2515/118, Rev.1, " Service Water System Operational Inspection (SWOPI)," 2/11/93
- 11. Inspection Reports Ginna, Monticello, Quad Cities, South Texas, St. Lucie, WNP-2, Watts Bar (Integrated Design Inspection)
- 12. IN 94-03: Deficiencies Identified During Service Water System Operational Performance Inspections," 1/11/94.
ASSESSMENT OUTLINE (Section numbers corresnond to NRC TI 2515/118) 03.01 Mechanical Systems Ertaineerina Desian Review and Configuration Control 03.01.a Review of design bases, calculations and analyses for SW to determirn functional requirements in accident and abnormal conditions. Assess design assumptions, bounding conditions, and models. Determine: a) design in accordance with licensing commitments & regulations b) thermal & hydraulic performance requirements c) drawing & procurement specifications consistent with design bases and engineering analyses SW reservoir flowpath Lake-to lake and make-up flowpath Transition of flowpaths
- NPSH (Ref. 8, Sec. 3.8)
As back up to CCW, AFW (Ref. 8, Sec. 3.12)
Room cooler capacity vs. all room heat loads (Ref.12, Sec.1.a) 03.01.b Determine consistency between drawings and: design, NRC requirements, licensing commitments.
03.01.c Search for discrepancies between operation and design documents.
HX & cooler design flow consistency with actual flow (Ref.12, Sec.1.b) 03.01.d Search for single active failure vulnerabilities and resulting impact on interfacing systems. Assess impact from failures in interfaci'ng systems.
ATTACHMENT 4 Page 98
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA A'ITACHMENT 4 ASSESSMENT PLAN (cont.)
Determine potential common mode failures from common structures or equipment.
Control systems, pumps, HXs (Ref. 8, Sec. 3.1)
- Common flowpaths (Ref. 8, Sec. 3.6)
Single failure as back up to CCW, AFW, etc. (Ref. 8, Sec. 3.7,3.12)
Common-mode failure from icing Motor and pump horsepower match: effect of UV, aging (Ref. 8, Sec.
3.17) 03.01.e Assess design to minimize biofouling and silting. Verify features to detect flow loss. Verify flow balancing is consistent w/ design assumptions for flow coefficients, rated component dPs, rated heat removal, HX fouling, and total system flow in various operating modes and worst case combinations of pump operation. Verify lack of pump run out at minimum pumps operational with worst case non safety-related loads. Evaluate minimum and maximum limits for valve positions and implementation into operational controls. (Ref. 8, Sec. 3.18)
SW reservoir flowpath !
Lake-to lake flowpath .
As back-up to CCW, AFW (Ref. 8, Sec. 3.12) '
- In-plant filters / screens 03.01.f Review design to mitigate flooding from SW leaks. (also Ref. 8, Sec. 3.19) 03,01.g Verify seismic qualification of safety-related sections and isolatability of non safety related sections.
SW pump house settling 03.01.h Review of a minimum of 3 SW modification packages for: 50.59, compatibility with design bases, revision of maintenance, operating, training, and PT procedures.
03.01.i Assessment of Action IV of GL 89-13.
03.01.j Assess monitoring of system performance degradation: trending, engineering evaluation, operability determinations.
03.01.k Determine consistency between alarm-actuation setpoints and design basis / assumptions.
03.02 Onerations 03.02.a Perform system walkdown for comparison with design drawings.
ATTACHMENT 4 Page 99
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA ATTACHMENT 4 ASSESSMENT PLAN (cont.)
03.02.b Review SW ops, ARPs, APs for consistency with design and implementability. Determine adequacy of flow instrumentation during accident. Review logs t. A termine adequacy of temperature and flow t monitoring.
SW reservoir flowpath Lake-to-lake flowpath Transition of flowpaths Control of cross ties (Ref. 8, Sec. 3.21) <
03.02.c Review operator training on SW for technical completeness, accuracy, and incorporation of modifications in manuals and lesson plans. Assess conduct of training on modifications.
03.02.d Review implementation of procedures for valve alignments in flowpaths to safety-related components for: compatibility of normal system alignment with accident conditions, methodology of throttle valve positioning, control of flow to heat exchangers due to changing temperatures.
SW reservoir flowpath Lake-to lake flowpath Transition of flowpaths Control of cross-ties 03.02.e Perform walk-through of ops and system drawings with operators and engineers. Use simulator,if available. Verify procedure performance and accessibility of equipment for normal and accident operation, availability and operability of special equipment, operator knowledge of equipment location and operation.
03.02.f Interview operators for technical knowledge of: system operation, TS surveillances, operability.
03.02.g Review local operation of equipment for: indicating equipment in accordance with ops, environmental conditions (temperatures, emergency lighting, steam) during accident conditions allow remote operation of equipment.
03.02.h Assess operation controls to prevent traveling screen clogging
- SW reservoir flowpath (Ref. 8, Sec. 3.3)
Lake to lake flowpath, Interaction with CW pumps, NPSH for SW pumps (Ref. 8, Sec. 3.3)
Common mode failure from icing 03.03 Mnintenance 03.03.a Perform walkdown for material condition.
ATTACHMENT 4 Page 100
i CORPORATE NUCLEAR SAFETY ASSESSMENT l NORTH ANNA POWER STATION SWSOPA ATTACHMENT 4 ASSESSMENT PLAN (cont.)
Common flowpaths (Ref. 8, Sec. 3.6) 03.03.b Witness some maintenance on SW system. Review WO package, observe QC involvement.
03.03.c Review maintenance procedures for technical content to perform intended task and have instructions to ID and evaluate deficiencies. Compare to vendor manuals. Verify vendor manuals are complete and updated.
03.03.d Assess maintenance program for corrective action for silting, biofouling, -
corrosion, erosion and failure of protective coating.
Debris intake (Ref. 8, Sec. 3.4)
E/C on and near flow, pressure, throttling valves, orifices (Ref. 8, Sec.
3.11)
SW chemical control program (Ref. 8, Sec. 3.10, 3.14)
- Protective coating monitoring (Ref. 8, Sec. 3.16) -
03.03.e Ensure maintenance is adequate for accident conditions. Review unavailability due to planned maintenance. :
03.03.f Review maintenance history for 2 3 years for recurring equipment problems and trends. Evaluate root cause analysis and corrective action ;
for adverse trends. Review several completed work packages for technical i adequacy, PMT, and demonstration of operability. ,
AOV and SOV malfunctions (Ref. 8, Sec. 3.22) 03.03.g Perform interviews with maintenance personnel to determine technical ;
knowledge of: component maintenance, setting limit switches, pump l coupling alignment, cleaning / replacing filters, circuit breaker maintenance FME controls '
03.03.h Review training of maintenance personnel and consistency with procedures.
Sufficient training for detection and evaluation of degraded components (Ref.12, Sec. 3.2).
03.03.i Assess inspection program to detect corrosion, erosion, protective coating failure, silting, and biofouling.
Debris intake (Ref. 8, Sec. 3.4)
Corrosion inspection areas: stagnant areas (incl. cross ties-Ref. 8, Sec.
3.121), dissimilar metals, etc. (Ref. 8, Sec. 3.10)
Microbiologically Induced Corrosion-(MIC) (Ref. 8, Sec. 3.13)
Biofouling monitoring (Ref. 8, Sec. 3.14)
ATTACHMENT 4 Page 101
CORPORATE NUCLEAR SAFETY ASSESSMENI' NORTH ANNA POWER STATION SWSOPA ATTACHMENT 4 ASSESSMENT PLAN (cont.)
Program scope covers NSR SW piping required for discharge or inlet flowpaths needed to perform design function (Ref.12, Sec. 3.2).
03.04 Survelliance and Testina !
1 03.04.a Technical adequacy and accuracy of TS surveillances and IST procedures i in last 2-3 years. Ensure design assumptions are reflected in test methods to demonstrate system performance.
03.04.b Review SW design and licensing basis to verify consistency with test acceptance criteria. Review system performance indicators to identify adequacy of test methods and/or frequency. Determine if surveillance procedures address SW system responses.
Verify inclusion in the IST program of: manual valves with specific reactor shutdown functions, manual and check valves used only i occasionally but have SR functions (i.e., backwash valves), valves i normally closed but may be needed to function in a DBA. (Ref.12, Sec. l 2.a)
Compliance with Section XI for pump testing, re. curves vs. reference values (Ref.12, Sec. 2.c). ;
Compliance with testing commitments of Action II of GL 89-13, i.e.,
room cooler testing, baseline minimum flow testing, basis for maintenance activities used to verify HX and cooler performance, i correction of test data taken under conditions that depart from design (Ref.12, Sec. 3.1).
03.04.c Review preoperational tests to determine demonstration of system capability and limitations. Determine if unacceptable operation of components was prevented. (Ref. 8, Sec. 3.18)
Determine consistency of preop tests with current system configuration and testing, i.e., isolation of NSR parts, testing performance under post-accident conditions, and establishing flow balance under most limiting case. (Ref.12, Sec. 2.d) 03.04.d Evaluate modifications reviewed by engineering team to ensure testing was adequate.
03.04.e Review IST records for pumps and valves for: technical adequacy of procedures, trending of results and recurrent failures. Completeness of IST program.
Test results are available for all required SWS components, i.e. control room HVAC. (Ref.12, Sec. 2.b)
ATTACllMENT 4 Page 102
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTil ANNA POWER STATION SWSOPA A'ITACHMENT 4 ASSESSMENT PLAN (cont.)
03.04.f Calibration and testing of instrumentation, valve stroke time, and installation of test equipment used for TS operability. Verify tolerances for instrument accuracy are acceptable. (Also Ref.12, Sec. 2.c) 03.04.g Witness PMT, surveillance and ISTs for SW.
03.04.h Review pts for safety related heat exchanger heat transfer capacity and trending of results.
03.04 i Determine component unavailability during last 2-3 years. Compare to IPE and determine any discrepancies.
03.04.j Verify components are tested to perform in accordance with their design basis.
03.04.k Review FTs to detect flow blockage from biofouling in other systems.
03.04.1 Review testing on one air-to-water HX for proper heat transfer. Examine air side for fouling.
03.05 Oun11tv Assurance and Corrective Actions 03.05.a Review SNSOC and MSRC minutes for past 6 months for SW items for discrepancies and unusual operability determinations.
03.05.b Review operational history: LERs, NPRDS,50.72 reports, enforcement actions, NCRs, TS operability determinations, WOs, and adverse test results or recurrent failures. Determine adequacy of RCEs.
Corrective actions reflected to other unit (Ref.12, Sec. 3.2).
03.05.c Compare results of team area assessments with quality verification programs in same areas to determine ability of those programs to identify same.
03.05.d Timeliness and technical adequacy of resolution of items from self-assessments. Review open items for tracking and closure.
03.05.e Evaluate interface between engineering / technical support and operations for resolution of operational problems.
ATTACliMENT 4 Page 103
l l
CORPORATE NUCLEAR SAFETY ASSESSMENT lR l NORTH ANNA POWER STATION SWSOPA j l
ATTACHMENT 5 !
I DOCUMENTS REVIEWED DESIGN '
Calculations 1
- ME-0294, Revision 0,8/29/91, Heat Transfer Capability Model for Component !
Cooling Heat Exchangers l
- ME-0420, Revision 1,6/1/94 (and Addendum A,6/14/94), Component cooling Heat Exchanger Retubing/ Replacement Study
- ME-0309, Revision 0,4/30/92, Base Case KYPIPE2 Service Water Flow Model for North Anna Power Station
- ME-0327, Revision 0,12/15/92, Service Water Operational Base Case KYPIPE2 Flow Model
- ME-0327, Revision 0, Addendum A,1/20/93, Service Water Operational Base Case ICfPIPE2 Flow Model i
- ME-327, Revision 0, Addendum B,3/8/94, Service Water Operational Base Case '
KYPIPE2 Flow Model
- CE-0667, Revision 0,5/3/89, Overpressurization of Service Water Piping
- CE-0860, " Service Water Lines Design Pressure Upgrade"
- ME-056, Revision 0,10/7/85, Spray System Water Hammer Analysis
- ME-0295, Revision 1,12/8/93, Verification of Adequacy of Piping Diameter to Control Room Chillers (Unit 2)
- ME-0224, Revision 0,4/17/89, Verification of Adequacy of Piping Diameter to Control Room Chillers (Unit 1)
SE-0011, Revision 0,11/22/91, Service Water Flow Balance for N 91-DR-1635
- ME-0322, Revision 0, dated 11/17/93, Service Water MOV Operating Torque Requirements
- ME-062,8/12/85 (with Addenda to 5/1/87), Reservoir Performance Analysis
- 14938.37 US(B)-259,5/28/87, LOCA Analysis for Revised Tech Spec
- 11715-ES-212-1, Revision 1,5/14/79, Extended Service Water Temperature Excursion '
- ME-0305,2/26/92, Service Water Pump NPSH
- 12.5.3.3.11,1/26/74, Service Water Spray Pond Thermal Performance
- N-179,1/23/74, Contract 11715, Plant Cooldown
- ME-086,3/5/86, Service Water Reservoir Volumes
- 163N,10/17/72, Contract 11715, SW Pump NPSH
- 12.5.3.3.19,12/28/76, Service Water Reservoir Surface Area vs. Water Elevation
- ME 0162, Revision 1, 7/8/88, Maximum Allowed Service Water Temp. to Recirc Spray HX after Wet Layup
- SM-691,5/17/89, UA Cale for Degraded SW
- ME-0173,4/14/89, Maximum Outside Tube Fouling Factor for the North Anna Unit I and 2 RSHXs
- ME-0200, Revision 0,11/04/88, Evaluation Of The Leakage Effects On The Service Water System ATTACHMENT 5 Page 104
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA F
ATTACHMENT 5 DOCUMENTS REVIEWED (cont.)
- ME-0164, Revision 1,6/12/89, Maximum Allowable Inside Tube Fouling Factor Of The Recirculation Spray Heat Exchangers To Assure Containment Depressurization
- 12050-432N, Revision 0,1/15/79, Determine The Minimum Service Water Flow To The Component Cooling Water (CCW) Heat Exchanger Required For Unit 2 With A LOCA In Unit 1 Licensinc Documents
- Portions of the NAPS UFSAR, including Sections 9.2.1,9.2.2,9.2.3,9.2.4, and 9.2.5
- Portions of SDBD-NAPS SW, Revision 01,12/31/93, Service Water System, North Anna Power Station, System Design Basis Document
- Portions of NUREG-0053, June 4,1976, Safety Evaluation Report (NRC)
- Technical Specifications,3/4 7.4 Service Water System
- Technical Specifications,3/4.7.5 Ultimate Heat Sink
- NUREG-0733, Analysis of Ultimate Heat Sink Spray Ponds Drawincs
- 11715-ESK-3A, Electrical Schematic Diag NAPS 1
- 11715 ESK-3H, Electrical Schematic Diag NAPS 1
- 11715-ESK-3W, Electrical Schematic Diag NAPS 1
- 11715 ESK-5A, Electrical Schematic Diag NAPS 1 11715 ESK 5AS, Electrical Schematic Diag NAPS 1
- 11715-ESK-5BC, Electrical Schematic Diag NAPS 1 11715-ESK-5BE, Electrical Schematic Diag NAPS 1
- 11715-ESK-10BG, Electrical Schematic Diag NAPS 1
- 11715-FB-3D-15, Revision 15, Yard - Water & Fire Protection Lines Sh - 4
- 11715-FB-40A, Revision 13, Sheet 1 of 3, Flow Diagram - Air Cond - Ch Wat Sys -
Sh 1
- 11715 FB-40A, Revision 13, Sheet 2 of 3, Flow Diagram - Air Cond - Ch Wat Sys -
Sh 2
- 11715-FB-40B Revision 11, Sheet 1 of 3, Flow Diagram - Air Cond - Cndnst Wat Sys - SH 1
- 11715 FB-40B, Revision 11, Sheet 2 of 3, Flow Diagram - Air Cond - Cndnst Wat Sys - SH 2
- 11715-FB-40C, Revision 15, Sheet 1 of 3, Flow Valve Operating No. Diag Air Cond Ch Wat Sys NAPS 1
- 11715-FB-40D, Revision 22, Sheet 1 of 3, Flow Valve Operating No. Diag Air Cond Ch Wat Sys NAPS 1
- 11715-FB-40D, Revision 26, Sheet 2 of 3, Flow Valve Operating No. Diag Air Cond Ch Wat Sys NAPS 1
- 11715 FB-40D, Revision 13, Sheet 3 of 3, Flow Valve Operating No. Diag Air Cond Ch Wat Sys NAPS 1 11715 FM-74A, Sheet 3, Flow Valve Operating No. Diag NAPS 1 ATTACHMENT 5 Pa;:e 105 l
I
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA ATTACHMENT 5 DOCUMENTS REVIEWED (cont.)
- 11715-FM 77A, Sheet 1 of 2, Flow Valve Operating No. Diag Circulating Water Sys NAPS 1
- 11715 FM-78A, Revision 48, Sheet 1 of 4, Flow Valve Operating No. Diag Service Water Sys NAPS 1
- 11715-FM-78A, Revision 27, Sheet 2 of 4, Flow Valve Operating No. Diag Service Water Sys NAPS 1- 11715 FM 78A, Revision 33, Sheet 3 of 4, Flow Valve Operating No. Diag Service Water Sys NAPS 1 ,
- 11715-FM-78A, Revision 64, Sheet 4 of 4, Flow Valve Operating No. Diag Service Water Sys NAPS 1
- 11715 FM-78B, Revision 28, Sheet 1 of 4, Flow Valve Operating No. Diag Service l Water Sys NAPS 1 l
- 11715 FM-78B, Revision 27, Sheet 3 of 4, Flow Valve Operating No. Diag Senice Water Sys NAPS 2
- 11715-FM-78C, Revision 35, Sheet 1 of 2, Flow Valve Operating No. Diag Service Water Sys NAPS 1
- 11715-FM-78C, Revision 28, Sheet 2 of 2, Flow Valve Operating No. Diag Service Water Sys NAPS 1
- 11715-FM-78F 8, Flow Valve Operating No. Diag Service Water Chemical Addition System NAPS 1
- 11715-FM 78G, Revision 15, Sheet 1 of 2, Flow Valve Operating No. Diag Service Water Sys NAPS 1
- 11715 FM-78G, Revision 11, Sheet 2 of 2, Flow Valve Operating No. Diag Service Water Sys NAPS 1
- 11715 FM-78H, Revision 8, Sheet 1 of 1, Flow Valve Operating No. Diag Service Water Sys NAPS 1
- 11715 FM-78J, Revision 0, Sheet 1 of 1, Flow Valve Operating No. Diag Service Water Sys NAPS 1
- 11715-FM-78K, Revision 0, Sheet 1 of 1, Flow Valve Operating No. Diag Service Water Sys NAPS 1
- 11715-FM-78L, Revision 7, Sheet 1 of 2, Flow Valve Operating No. Diag SW Chem Add Sys NAPS 1
- 11715 FM 78L, Revision 5, Sheet 2 of 2, Flow Valve Operating No. Diag SW Chem Add Sys NAPS 1
- 11715-FM-79C, Revision 15, Sheet 3 of 5, Flow Valve Operating No. Diag NAPS 1
- 11715-FM-79D, Sheet 4, Flow Valve Operating No. Diag NAPS 1
- 11715-FM 87C, Sheet 1, Flow Valve Operating No. Diag NAPS 1
- 11715 FM-105B, Sheet 1, Flow Valve Operating No. Diag NAPS 1 11715 FM 105B, Sheet 2, Flow Valve Operating No. Diag NAPS 1
- N8546 3-AM 100, Revision 0, Sheet 1 of 1, Corr RT Sens PRB Inst - SW Corr Rate Monit - NAPS 1&2 N8546-3-AM-101, Revision 0, Sheet 1 of 1, Corr Rate Coup Instal - SW Corr Rate Monit - NAPS 1&2
- N8546-3 AM-103, Revision 0, Sheet 1 of 1, Retr Coupon Hidr Assy - SW Corr Rate Monit - NAPS 1&2
- N8546-3-AE-110, Revision 0, Sheet 1 of 1, Corr Rt Mon Blok Diag - SW Corr Rate Monit - NAPS 1&2 '
ATTACHMENT 5 Page 106
1 CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA ATTACHMENT 5 DOCUMENTS REVIEWED (cont.)
- N8545-3 M-108, Revision 0, Sheet 1 of 1, Corr Rat Monitoring Station
- N8499-1 M-106, Revision 0, Sheet 1 of 2, Corr Rate Monitor for 24"-WS-26-151-Q3 Pipe
- N84991-M-106, Revision 0, Sheet 2 of 2, Corr Rate Monitor for 24"-WS-26-151-Q3 Pipe DCPs. Technical Reoorts. Soecifications
- Specification Data Sheets NAS-96, " Component Cooling Heat Exchanger"
- Technical Report No. ME-0051, Revision 0,12/13/91," Component Cooling Heat Exchanger Performance Testing, NRC Generic Letter 89-13
- Technical Report No. ME 0074, Revision 0, 7/1/93, " Component Cooling Heat Exchanger Performance Testing of 2 CC-E-1A"
- Technical Report No. ME-0026," Service Water Single Failure Review North Anna Power Station," 10/25/90
- Technical Report No. ME-0029, Revision 0, " Control Room Chiller Equipment .
Performance Test"
- Report: "SW Reservoir and Spray System Performance Testing & Evaluation,"
2/79, Ford, Bacon, & Davis, Utah
- DC-85-48-3, " Service Water Chemical Addition System / North Anna /1&2,"
Engineering Review and Safety Analysis (Supplement)
- DCP 86-02-01,10/8/86, Core Uprating
- Specification NAS-98
- NAS 160, " Specification for Recirculation Spray Coolers For North Anna Power Station"
- Type I Draft Report, " Internal Flooding Of Power Plant Buildings, North Anna Power Station Units 1 & 2"
- Type I Final Report, "Outside Containment Flooding Protection," 12/14/89 Procedures
- VPAP-0301, " Design Change Process," Revision 4,8/04/94
- 0 GOP-4, " Cold Weather Operations," Revision 6 GOP-4.2, " Extreme Cold Weather Operations," Revision 1 PT-75.6, " Service Water System Flow Balance, " completed 4/07/93 PT-75.6, " Service Water System Flow Balance, " completed 10/21/93 LOG-4, "CRO Surveillance Sheets," Revision 22 LOG-4A, "CRO Surveillance Sheets (Modes 5&6)," Revision 14 LOG-4, "CRO Surveillance Sheets," Revision 26 LOG-4A, "CRO Surveillance Sheets (Modes 5&6)," Revision 16 OP-21.1, " Containment Ventilation," Revision 20 17-57.4 17 66.3 Miscellaneous ATTACllMENT 5 Page 107
1 CORPORATE NUCLEAR SAFETY ASSESSMENT l I NORTH ANNA POWER STATION SWSOPA A'1TACHMENT 5 DOCUMENTS REVIEWED (cont.)
- Memorandum POW-36-82,5/4/87, R. L. Rasnic to J. A. Stall, Service Water Reservoir Improvements, North Anna Units 1 and 2, Spray / Bypass System Operating Guidelines
- Serial No.89-572,1/29/90, VEPCO letter to NRC, Generic Letter 89-13: Service Water System Problems Affecting Safety-Related Equipment
- Minutes of SDBD Open Item Review Meeting for SDBD-NAPS-T V, " Service Water System, North Anna Power Station,"(Undated)
- Safety Evaluation 94-SE-0T-034,5/26/94, "CCHX Tube Plugging"
- Justification for Continued Operation, JCO 92-05, Revision 1,6/7/94 " Evaluation of Service Water Piping Integrity for Concrete Encased Piping"
- Johnston Pump Co Service Water Pump Curves,1,2 SW-P-1A,B; 1,2 SW-P-4.
- Doc.#46 MDS-2030S-1, Revision 1, NUREG-0733 Analysis , Computer Program Documentation File
- Doc.#60 MDS-3258-1, Revision 1, E&C Mech Eng, User's Guide Supplement to NUREG-0733 Analysis
- Individual Plant Examination, HNUS Project 0581, Internal Flooding Analysis, Interim Report, February 28,1992
- Conceptual Engineering Report, IPE Flooding Analysis, Evaluation Of Proposed Plant Modifications, North Anna Power Station, Revision 1, May 7,1992
- 1991 Performance Evaluation Report, North Anna Power Station Main Dam, Reservoir, Cooling Lagoons, Dikes And Canals, And Initial Inservice Inspection Flood Protection Dike, February 1992
- North Anna Setpoint Document OPERATIONS Drawincs (NOTE: The major SW drawings, the 11715-FM-78 series, are listed under design, but were also reviewed in detail for this section. Other drawings not previously mentioned are listed below.)
- 11715-FB-40A-13, Air Conditioning Chilled Water Systems-Sh-1
- 11715 FB-40A-13, Air Conditioning Chilled Water Systems-Sh-2
- 11715 FB-40B-11, Air Conditioning Condenser Water Systems Sheet 1
- 11715-FB-40B-11, Air Conditioning Condenser Water Systems Sheet 2
- 11715-FB-40D-22, Air Conditioning Condenser Water Sys, Sheet 1
- 11715 FB 40D 26, Air Conditioning Condenser Water Sys, Sheet 2
- 11715-FB-40D-13, Air Conditioning Condenser Water Sys, Sheet 3
- 11715-FC-19B-13 Sheet 2, Main Stm Valve, Quench Spray Pump HSGS &
Safeguards Area
- 11715-FC-19D 6
- 11715 FC-19M-6, Main Steam Valve Housing Plans
- 11715 FC-19Q 5 Main Steam Valve Housing East and West Wall
- 12050-FC-19A-9, Main Stm Valve, Quench Spray Pump HSGS & Safeguards Area ATTACHMENT 5 Page 108
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CORPORATE NUCLEAR SAFETY ASSESSMENT {
l NORTH ANNA POWER STATION SWSOPA ATTACHMENT 5 DOCUMENTS REVIEWED (cont.) ;
- 12050 FC-19B-13, Sheet 2, Main Stm Valve, Quench Spray Pump HSGS & 1 Safeguards Area
- 12050-FC-19C-6, Main Stm Valve, Quench Spray Pump & Safeguards Area
- 12050 FC-19D-7, Main Stm Valve, Quench Spray Pump HSGS & Safeguards Area
- 12050 FC-19M 9, Main Steam Valve Housing Plan
- 12050-FC-19P-8, Main Steam Valve Housing
- 12050 FC-19Q-7, Main Steam Valve Housing
- 12050 FM-1C-11, Sht 3, Mach Loc - Reactor Cont ,
- 12050-FP-7C Rev.10, Yard Piping- North Reactor Containment '
- 12050-FP-7D Rev.12, Yard Piping- North Reactor Containment
- 12050-FP-7G-10, Yard Piping- North Reactor Containment
- 12050 FP-7H-9, Yard Piping North Reactor Containment Procedures reviewed in detail OP-49.1, Rev. 3, Service Water System Normal Operation OP-49.1A, Rev.17, Valve Checkoff-Service Water OP-49.2, Rev. 2, Service Water System Lake-to-Lake Operation
-0 OP-49.3, Rev. 2, Service Water Reservoir Makeup OP-49.4, Rev. 5, Shifting Service Water Components (Pumps and Sprays)
-0 OP-49.6, Rev. 3 P1, Service Water System Throttling Alignment OP-21.1, Rev. 20, Containment Ventilation OP-21.6, Rev.14, Main Control and Relay Room Air Conditioning OP-21.11, Rev. O, Bearing Cooling Supply to the Unit 1 Control Room Chillers
-1 OP-48.1, Rev. 9, Circulating Water Screenwash System OP-21.1, Rev.17, Containment Ventilation AP-12, Rev. 5, Loss of Service Water AP-27, Rev. 2, Malfunction of Spent Fuel Pit System AP-39.1, Rev. 2, Turbine Building Flooding AP-39.2, Rev.1, Auxiliary Building Flooding AP-40, Rev. 3, Abnormal Levelin North Anna Reservoir (Lake) AP-40.2, Rev 2., Dam Failure Assessment and Notification AP-41, Rev. 9, Severe Weather Conditions
-0 AP-47, Rev 3, Unit Operation During Opposite Unit Emergency AP-5, Rev. 5, Unit 1 Radiation Monitoring System AP-5.1, Rev. 5, Common Unit Radiation Monitoring System AP-22.5, Rev. 3, Loss of Emergency Condensate Storage Tank CN-TK-1 AP-35, Rev. 5, Loss of Containment Air Recirculation Cooling
-0 PT-75.1, Rev. 6, Service Water System-Valves (Monthly)
-1 PT-75.2B, Rev. 26 P1, Service Water Pump (1-SW-P-1B) Quarterly Test PT-75.4, Rev. 7, Service Water Screen Wash Pump (1-SW-P-2) PT-62.2.1, Rev. 8, RSHX SW Inleakage PT-62.2.1A, Rev. 2, RSHX SW Inleakage PT-75.5, Rev. 20, Auxiliary Service Water Pump (2-SW-P-4) Test ATTACllMENT 5 Page 109
CORPORATE NUCLEAR SAFETY ASSESSAf ENT NORTH ANNA POWER STATION SWSOPA ATTACHMENT 5 DOCUMENTS REVIEWED (cont.)
-0 FCA-9, Rev. 2, Service Water Pump House Fire MOP-49.09, Rev. O, Removing Service Water No. 2 (B) Supply and No. 3 (B)
Return Headers from Service and Returning to Senice MOP-49.11 Rev. O, Service Water Flooding in Auxiliary Building 0-MOP-49.12, Rev. 2-P3, Service Water System Chemical Treatment 0-MOP-49.30, Rev.1, Draining SW form CC Heat Exchangers MOP-49.01, Rev. 7,1-SW-P-1A, Service Water Pump MOP 50.31, Rev. 3, Component Cooling Heat Exchanger 1-CC-P-1B MOP-50.31, Rev. 4, Component Cooling Heat Exchanger 2-CC-P-1B
-0 LOG 6A, Rev. 2, Backboards Tour LOG-6E, Rev.12, Outside Log Annunciator Response Procedures:
1A-C4,Rev.34 IB-C8,Rev.31 1E-C5,Rev.16 1E C6, Rev.16 1E-C7,Rev.16 1E-F5,Rev.16 IE F6, Rev.16 IG-A1, Rev. 23 IJ-B3, Rev. 22 IJ B4, Rev. 22 1
1J D3, Rev. 22 1 1J-D4, Rev. 22 IJ D5, Rev. 22 !
IJ-E3, Rev. 22 IJ-E4, Rev. 22 IJ-E5, Rev. 22 IJ-E6, Rev. 22 1J E7, Rev. 22 1 IJ F8, Rev. 22 !
IJ-G7, Rev. 22 l 1J-H3, Rev. 22 IJ-H6, Rev. 22 1K D2, Rev. 23 IK-D4,Rev.23 IK-D5, Rev. 23 IK-E3,Rev.23 IK-E4, Rev. 23 1K-F1,Rev.23 IK-F2,Rev.23 IK-F3,Rev.23 '
1K G5, Rev. 23 ATTACHMENT 5 Page 110
CORPORATE NUCLEAR SAFETY ASSESW1bNT NORTH ANNA POWER STATION SWSOPA ATTACHMENT 5 DOCUMENTS REVIEWED (cont.)
MAINTENANCE Work Orders
- 5900134658: 2-SW-MOV-203C
- 5900142331: 02-SW-MOV-201A
- 5900123310: 01-SW-MOV-104B
- 5900122113: 01-SW. MOV-104B
- 5900121242: 01-SW-MOV-104B
- 5900121593: 01-SW MOV-104B
- 5900108479: 01-SW-MOV-104B
- 5900140350: 02-SW-MOV 201A
- 5900126973: 02-SW-MOV-201A
- 5900130201: 02-SW-MOV-201A
- 5900125849: 02-SW-MOV-201A -
- 5900142320: 02-SW-MOV-2030
- 5900145810: 02-SW-MOV-2030
- 5900152463: 01-SW-63
- 5900154463: 01-SW-22
- 5900152458: 01-SW-343 5900116895: 01-HV-E-4A
- 5900130025: 01 HV-E-4A
- 5900131558: 02-HV-E-4C
- 5900150523: 02-HV-E-4C
- 5900132558: 02-HV-E-4C
- 5900157875: 01-SW-1214
- 5900142996: 02-SW-P-2
- 5900159297: 01-SW-REJ 6A
- 5900159298: 01-SW-REJ-6B
- 5900159299: 01-SW-REJ-60
- 5900042502: 02-SW-P-1B
- 5900042503: 01-SW-P-1A
- 5900042504: 01-SW-P-2
- 5900061108: 01-SW-P-1A
- 5900154685: 01-SW-S-1A
- 5900161216: 01-SW-S-1 A
- 00275159 01: 01-SW-1125
- 00267594 01:1-SW-55
- 00280055 01,03: 1-CC-E-1A
- 00280055 02: 01-SW-226
- 00288658 01: 1-CC E-1B
- 00284257 01: 2 CC-E-1A
- 00286905 01: 2-CC-E-1B
- 00287816 01: 1-SW-P-1B (motor) ~
- 00274290 01: 01-SW-S-1 A ATTACHMENT 5 Page 111
l CORPORATE NUCLEAR SAFETY ASSESSMENT l NORTH ANNA POWER STATION SWSOPA ATTACHMENT 5 !
DOCUMENTS REVIEWED (cont.) !
- 00287818 01: 01-SW-S-1A
- 00289820 01: 01-SW-RV-111A
- 00262614 01: 01-SW-RV-111B
- 00259612 01,02: 02-SW-RV-200A
- 00259613 01,02: 02-SW RV-200B )
1 Procedures
- 0 ECM-1401-03, Rev.1, General Maintenance of Electric Motors l ECM-1501-02, Rev. O, Disconnection and Reconnection of Motor Operated Valves 1
- 0 ECM-1501-03, Rev. O, Predictive Analysis of Motor-Operated Butterfly Valves
- 0 ECM 1502-05, Rev. O, Inspection and Repair of Limitorque Valve Control Units, 1 Types SMB-000 and SMB-00, with HBC Actuators for Butterfly Valves !
ECM 1506-01, Rev. O, Replacement of Limitorque Torque Switches Identified in :
10 CFR Part 21 Notification From Limitorque
- 0 EPM-1412-01, Rev. 2, General Inspection and Testing of Electric Motors
- 0 EPM-2802 01, Rev. O, Disconnecting and Reconnecting Electrical Equipment i EPM-2803-01, Rev. O, Disassembly and Reinstallation of Raychem Splices '
EPM-1815-02, Rev. O, Protective Relay Maintenance for Breaker 15J5 Service Water Pump 1-SW-P-1B EPM 1815-03, Rev. O, Protective Relay Maintenance for Breaker 15H4, Service l Water Pump 1 SW-P-4 i
- 0 MCM-0115-01, Rev.1, Repair of the Main and Auxiliary Service Water Pumps ' MCM-0116-01, Rev. O, Repair of the Service Water Screen Wash Pump MCM-040012, Rev. 2, Disassembly, Inspection, and Repair of Non-Safety Related Check Valves in General
- 0 MCM-0400-14, Rev. 2, Repair of Safety Related and Non-Safety Related Safety and Relief Valves in General MCM-0400-30, Rev. O, Removal, as Left Testing, and Installation ofIST Safety and Relief Valves MCM 0400 31, Rev.1, Removal, Testing, Repair, and Installation of Safety and Relief Valves (for Replacement of non-IST Valves) MCM-0400-32, Rev. O, Disassembly, Inspection, and Repair of Safety Related Check Valves in General
- 0 MCM-0404-01, Rev.1, Disassembly, Repair and Reassembly of Service Water Butterfly Valves MCM-0431-01, Rev.1, Mission Duo-Check and C&S Dual Plate Check Valve Repair MCM 0437-01, Rev. 3 (PI), Repair of Hirata and Pacific Swing Check Valves
- 0 MCM 0463 01, Rev.1, Vogt Piston Check Valve Repair MCM-0537-01, Rev. 3, Pacific Swing Check Valve Repair ;
- 0 MCM-1004 01, Rev.1, Removal and Installation of Rubber Expansion Joints,24- '
inch and Smaller
- 0 MCM 1304 01, Rev. 2, Mming Missile Shield Blocks in the Turbine Building, Protected Area, and Yard 0-MOP-49.12, Rev. 2 (P3), Service Water System Chemical Treatment ATTACllMENT 5 Page 112
CORPORATE NUCLEAR SAFETY ASSESSMENT l NORTH ANNA POWER STATION SWSOPA ll 1
ATTACHMENT 5 i DOCUMENTS REVIEWED (cont.)
MPM-0103-01, Rev. 3, Preventive Maintenance of Charging /High-Head Safety Injection Pumps MPM-0103 01, Rev. 4, Preventive Maintenance of Charging /High-Head Safety Injection Pumps MPM-1004-01, Rev. 2, Preventive Maintenance Inspection of Rubber Expansion Joints MOP-49.31, Rev. 4, Draining Recire Spray Heat Exchangers (Service Water Side)
- MEMP-C-MOV-6, Mechanical Electrical Procedure for Safety-Related Motor Operated Valve Repair and Inspection in General,06-30-88
- MMP-P-MR-1, Periodic Disassembly, Inspection, and Repair of the Control Room l Air Conditioning Chillers, dated 10-27-88 MMP-P-MR-1, Rev. 4, Periodic Disassembly, Inspection, and Repair of the Control Room Air Conditioning Chillers
- MMP-P-SW NSR-1, Rev. O, Preventive Maintenance Inspection of Service Water Reservoir Instrument Air Compressor .
- M-10-MOV/R 6(EQ), Limitorque Motor Operated Valve Gear Case and Valve Stem Inspection, dated 07-19-90
- M-10-MR/A-8, Rev.1, Inspection and Service of Control Room Chiller
- M-10-TS/SA-1, Lubrication and Service of Service Water Traveling Screens
- E-14-MOV/R-4(EQ), Limitorque Motor Operated Valve Inspection and Service, dated 07-19-90
- E-24-MOV/R-4(EQ), Rev. O, Limitorque Motor Operated Valve Inspection PT-75.14, Rev.1, Service Water Wall Thickness Monitoring PT-75.15, Rev.1, Generic Letter 89-13 Service Water System Testing Requirements Coordination
- 1 PT-66.3, Rev.17, Containment Depressurization Actuation Functional Test
- 1 PT-75.4, Rev. 7, Service Water Screen Wash Pmnp (1-SW-P-2)
PT-111, Service Water Reservoir Pump House and Dike Settlement Monitoring, dated 10-4-84 PT-48, Rev. 5, Visual Inspection of ASME XI Class 1 Pressure Boundary Components PT-75.4, Rev. 8, Service Water Screen Wash Pump (2-SW-P-2)
Miscellaneous
- Individual Plant Examination, North Anna Power Station Units 1 & 2, Interim Report, February 28,1992
- Summary printout of Work Orders on Service Water and Control Room Chillers from 1989 to 1994
- Materials Engineering Laboratory Report, NESML-Q-116, CCHX tube failure analysis, Feb. 10,1994
- Listing of personnel and completed JPMs .
Listing of course schedules for initial and continuing training for electricians and mechanics.
ATTACHMENT 5 Page 113
CORPORATE NUCLEAR SAFETY ASSESSMENT NOftTH ANNA POWEit STATION SWSOPA ATTACHMENT 5 DOCUMENTS REVIEWED (cont.)
TESTING Licensinc Documents
- GL 8913 and Virginia Power submittals.
- Notice of Violation dated March 2,1992 -- North Anna -- NRC inspection Report Nos. 50 338/90-02 and 50-339/90-02
- Service Water Technical Specifications
- NRC SER to IST Program Rev. 6
- IST Program Submittal to NRC including Relief Requests Imnlementinc Procedures
- NASES-6.18 Rev. O Controlling Procedure for the Inservice Testing (IST) Pump Engineer. Issued January 31,1994.
- NASES-6.17 Rev. O Controlling Procedure for the Inservice Testing (IST) Valve Engineer. Issued December 1,1993.
- VPAP-0811, Service Water System Inspection and Maintenance Program.
Effective date 2/7/94.
Periodic Tests
- 1,2-PT-77.13A,B,C, Control Room Chiller Equipment Performance Tests (1,2-HV-E-4A,B,C); including endneering analysis.
- 1,2-PT-77.11A,B,C Control Room Chiller Pump and Valve Test PT-75.6," Service Water System Flow Balance" PT-75.7, " Service Water Reservoir - Ground Water Level" PT-214.12,"SW Valve Position Indication" PT-115, " Survey of Settlement Monitoring Points" PT-75.6.2, " Service Water Pump House Drain System - Flow Rate and Clarity" PT-75.7," Service Water Reservoir Ground Water Level" PT-57.4, ", " Safety Injection Functional Test" PT-66.3 Containment Depressurization Actuation Functional Test" PT-75.8, " Service Water Reservoir Loss Monitoring Procedure" PT-39.4," Triaxial Peak Acceleration Recorder Calibration" PT-39.5, " Triaxial Response Test Spectrum Recorder Channel Check" PT-39.6," Triaxial Response Spectrum Recorder Calibration" Various Valve Inservice Inspection Procedures for Supply Check Valves to Charging Pump Lube Oil, Seal, and Gear Box Coolers.
- Service Water Pump Quarterly Tests
- Auxiliary Service Water Pump Quarterly Tests
- Service Water Pump Head Curve Verification including analysis of degradation
- Valve Inservice Inspection (stroke testing of MOVs)
- Performance Tests of Control Room Chillers Desien Chances. Encineerine Work Reouests (EWRt Technical Renorts' A'ITACIIMENT 5 Page 114
CORPORATE NUCLEAR SAFETY ASSESSMENT NORTH ANNA POWER STATION SWSOPA ATTACHMENT 5 DOCUMENTS REVIEWED (cont.) 123 SW Strainer Replacement
- 92 266 Replacement of SW pump
- 84 31 Spray Array 04 SW 4" Control Room Chiller Piping Reroute
- 84 385 MOV Stroke Time Requirements dated 7/12/84.
- Technical Report No. ME-0051, Rev. O Dated December 13,1991.
Work Orders
- 00279910-01: 1-SW 22
- 00275446-01: 1-SW-MOV-102B-VALVOP
- 00276014-01: 1-SW-MOV-108A-VALVOP
- 00275857-01: 1-SW-MOV-108B-VALVOP
- 00288000-01: 1-SW-P-1A-MOTOR
- 00290518 01: 1-SW-TCV-102B-VALVE
- 00265098-14: 1-SW-MOV-217-VALVOP
- 00279749 01: 2-SW-5-1B-MOTOR
- 00285944 01: 2-SW-TCV-202B-VALVE Miscellaneous
- LER 89-008 SW Flow to RSHX Less Than Design Assumptions
- LER 90 012 SW System Operated In an Unanalyzed Condition Causing Possible Low Flow To Recirculating Spray Heat Exchangers Due to Personnel Error
- LER 88-024 SW Flow Not Within UFSAR Assumptions
- Specification NAS-160 Recirculation Spray Coolers
- Specification NAS-96 Component Cooling Water Heat Exchangers
- Calculation ME-0317 and ME-322 -- MOV Thrust Calculations
- Monthly Report ofIST Pumps and Valves Program From January 1992 to 1994.
CORRECTIVE ACTION Licensee Event Reoorts LER 89-005 LER 89 017 LER 89-019 LER 89 021 LER 89 023 i LER 89 024 '
LER 89-029 LER 89-030 LER 89-031 LER 89-034 ATTACHMENT 5 Page 115 l
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l CORPORATE NUCLEAR SAFETY ASSESSMENT l
NORTH ANNA POWER STATION SWSOPA l ATTACHMENT 5 l DOCUMENTS REVIEWED (cont.)
LER 90 001 '
LER 90-012 LER 90 014 LER 91-002 LER 91-005 LER 91-011 LER 91-016 LER 91-019 LER 92-002 LER 92-003 LER 92-009 LER 93 001 LER 93 006 LER 94-003 Miscellaneous
- Deviation Reports
- QATT entries
- IOD Reports
- Cold Weather Protection Assessments (1992 and 1993) ,
- OPS /Maint. Interface and Return to Service of Equipment Assessment
- Low Level Materiel Condition Assessment
- Review of CCfECs i
ATTACHMENT 5 Page 116 l
CORPORATE NUCLEAR SAFETY ASSESSMENT ;
NORTH ANNA POWER STATION SWSOPA ATTACHMENT 6 SWSOPA TEAM MEMBER RESUMES 1
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ATTACHMENT G Page 117
J. MICHAEL SURFACE S:nior St:ff Engin=r, Corportt3 Nucl:ar Saf ty Team Leader - North Anna Service Water System Assessment EDUCATION: I l
M. S., Chemical Engineering, University of Virginia,1982.
B. S., Chemistry, College of William and Mary,1975.
LICENSES / SPECIAL TRAINING:
INPO HPES Evaluator Training,1990 EXPERIENCE:
1987 - 1994 VIRGINIA POWER COMPANY Senior Staff Enaineer. Corocrate Nuclear Safety (6/89 Present). Conduct plant assessments and review documents in areas important to nuclear safety. (See Assessment Experience below.) Also perform Independent Reviews of safety evaluations in areas of mechanical engineering, metallurgy, chemistry &
radiochemistry, and radiological safety. Prepare Industry Operating Experience analysis reports.
System Chemist. Corocrate Office (1/87-6/89). Chemistry support to fossil and hydro I
stations. Chemical equipment specialist. Responsible for conduct of chemical cleaning, procedure development, fuel and tube oil sampling and analysis, boiler /feedwater chemistry, environment chemistry liaison, equipment design review, training.
1982-1986 GEORGIA POWER COMPANY, PLANT VOGTLE Lead Test Suoervisor (10/84-12/86). Start-up Engineering Supervisor for 15 chemical and radwaste system engineers. Component testing, flushing, initial start-up and preoperational testing supervision. System responsibilities included: 1) Make-up demineralizer,2) Make-up water storage and treatment,3) Condensate polishers,4)
Chemical addition systems (including service water and circulating water chemical injection),5) Waste gas,6) Boron and waste evaporators,7) Backflushable filters,8)
Sampling systems,9) Advanced radwaste processing equipment.
Grouc Leader. Chemical Enaineerina (10/8210/84). Lead on Chemical Engineering staff. Performed procedure preparation and review, FSAR reviews, design review, configuration control, SGOG activities, regulatory issues affecting Chemistry and radwaste.
Plant Engineer (2/82-10/82). Staff / start up engineer for chemistry related activities.
Procedure index/ schedule preparation, design reviews, general training.
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J. MICHAEL SURFACE S:nior Stcff Engineer, Corpor2t2 Nucirr Saf ty (cont.)
1975-1980 VIRGINIA ELECTRIC AND POWER COMPANY Staff Chemist. General Office (7/76 6/80). Chemistry support activities including pure water production, chemical cleaning, boiler /feedwater treatment, operator training, environment, corrosion and scale control. Oversight of chemistry program at North Anna and Surry. Chemistry contact on service water corrosion / deposition problem at North Anna.
Chemist. Yorktown Power Station (8/75 7/76). Boiler / feedwater and fuel analyses.
APPLICABLE ASSESSMENT EXPERIENCE:
Team Leader:
- Service Water System Performance Assessment (SWSOPA), Surry ,4/94
- Electrical Distribution System Functional Assessment (EDSFA), Surry,5/92.
- EDSFA follow-ups (2), Surry,12/92 & 1/94.
Event review of containment chiller failing to trip, Surry,12/92.
- Missed surveillances, Surry and North Anna,7/91. .
Participant:
- NUMARC Shutdown issues, Surry and North Anna,1/93.
- EDSFA, North Anna,4/91.
- Procedure assessment, Surry and North Anna,11/92
- Switchyard activities, Surry and North Anna,4/92
- IPE assessment, Surry, 3/92.
- Decay Heat / LOOP Vulnerablity Assessment, Surry and North Anna,6/90.
- Commitment management, Surry and North Anna,1/90 Utility Exchange Assessment:
- Mzintenance assessment, Zion (Com. Ed.),3/90.
FRANCIS T.TERMINELLA Senior St ff Engineer, Corporito Nucl=r Safety Assistant Team Leader - North Anna Service Water System Assessment EDUCATION:
M. S., Mechanical Engineering, Virginia Polytechnic Institute & SU,1975 B. S., Mechanical Engineering, Virginia Polytechnic Institute & SU,1973 LICENSES / SPECIAL TRAINING:
Shift Technical Advisor, North Anna Power Station,1980 Senior Reactor Operator's License, North Anna Power Station,1980 MORT training at INPO EXPERIENCE:
1990- 1994 VIRGINIA POWER COMPANY CORPORATE OFFICES Senior Staff Enoineer. Corocrate Nuclear Safety (9/90 Present). Responsible for performing assessments, (See Assessment Experience below), preparing Industry Operating Experience analysis reports and independent reviews of safety analyses.
1973 - 1990 NORTH ANNA POWER STATION, VIRGINIA POWER Suoervisor - Quality Assurance (8/84 9/90). Technical Advisor to the OA Manager. ,
Supervised QA personnel, implemented independent assessments of Operations and Maintenance Department activities.
Enaineerina Suoervisor (1/82-7/84). Responsible for supervision of all STAS.
Enaineerina Suoervisor (11/80-12/81). Responsible for supervision of design engineers.
Shift Technical Advisor (5/80-10/80). On-shift STA duties.
SRO Trainee (2/79-3/80). SRO license obtained in 1980 and maintained active until 1986.
E Technical Enaineer (1/79). Responsible for Unit 2 initial Hot Functional testing.
SRG l rainee (10/78-12/78). Assigned to an operating shift to learn systems.
Performance Enaineer (10/77-10/78). Reviewed completed periodic tests. Wrote Periodic Tests to ensure that surveillance requirements of Technical Specifications were performed. Set up Pump Performance Monitoring Program in accordance with ASME XI.
, -. , - - - . . - , _ - - - - , _. -rys e ea s 'PA sw _ __
FRANCIS T.TERMINELLA S nior St:ff Engineer, Corporats Nuclair S fity (cont.)
Start-uo Enaineer (7/73-10/77). Wrote preoperational and periodic test procedures.
Verified flushing of piping systems. Performed system checkouts of Unit 1 equipment prior to initial release of equipment. Conducted pre-operational testing.
APPLICABLE ASSESSMENT EXPERIENCE:
Team Leader:
- Service Water System Operathnal Performance Assessment (SWSOPA) Assistant Team Leadar, Surry,4/94.
- Shutdown Management Follow-Up, North Anna and Surry,3/93.
- Shutdown Management Assessment per NUMARC 91-06, North Anna and Surry, 2/92.
- Safety System FunctionalInspection (SSFI) Readiness Assessments at North Anna, Auxiliary Feedwater System,1987 and Instrument Air System,1988.
Participant:
- Electrical Distribution System Functional Assessment (EDSFA), Surry,5/92.
- EDSFA follow-ups (2), Surry,12/92 & 1/94.
- Operations Events Assessment, North Anna,5/92.
- Component mispositions/ transpositions, Surry and North Anna,10/91.
- Configuration Control Assessment, Surry and North Anna,5/91.
- EDSFA, North Anna,4/91.
- Lubrication Control Assessment, Surry,3/91.
Utility Exchange Assessments:
- Operations activities assessment, Shearon Harris (CP&L),8/93.
- Maintenance assessment, Zion (Commonwealth Edison),3/90.
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JOHN E. LEWIS S:nior Staff Engineer, Corpor2te Nucl :r S:f;ty EDUCATION:
B. S., Mechanical Engineering, University of Manchester,1975.
Master of Business Administration, University of Richmond,1991.
LICENSES / SPECIAL TRAINING:
Shift Technical Advisor,1979 EXPERIENCE:
1987 - Present VIRGINIA POWER COMPANY Senior Staff Enaineer. Corocrate Nuclear Safety (1/94-Present). Responsible for preparing industry Operating Experience analysis reports.
System Enaineer (6/85-12/93). Provided technical advice and assistance to plant management to support operation and maintenance of plant systems. Systems included Circulating Water, Reactor Coolant, instrument Air, Auxiliary and Main Feedwater, and Residual Heat Removal. Participated in Service Water flow balancing and MOV set-up. Reviewed plant modifications for compliance with standards and operational requirements. Provided technical support for Surry life extension studies.
Participated in SSFl's (See Assessment Experience below) and performed design basis reconstitution review of systems. Provided direction for revision of maintenance operational testing and plant performance programs; co ordinated daily activities of the System Engineering department.
1982 -1985 STONE & WEBSTER ENGINEERING CORPORATION Test Enaineer. Directed preparation and review of preliminary and pre-operational mechanical and electrical test procedures for power plants. Supervised preoperational mechanical and electrical testing at Shoreham and Nine Mile Point 2. Performed Post Modification Testing at Salem. Condui ad assessments of operability and readiness of plant systems prior to plant startup.
1978 - 1982 CONSOLIDATED EDISON OF NEW YORK Shift Technical Advisor. Indian Point Unit 2 (12/79-4/82). Provided technical and analytical suppcrt to management and the Shift Supervisor in evaluating non routine l plant conditions. Performed safety evaluations and coordinated and directed activities 1 of plant personnel involved with plant maintenance and operation during refueling i outages. !
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JOHN E. LEWIS Ssnior Staff Enginsar, Corporata Nuclar Safsty (cont.)
Cadet Enaineer. Arthur Kill Generatino Station (9/78-12/79). Supervised planning and scheduling of maintenance of instrumentation and controls (boiler and turbine generator). Monitored and evaluated performance of boiler, turbine generator, pumps,
- and fans. Outage coordinator during outages.
1974 - 1976 GUYANA BAUXITE COMPANY Cadet Enaineer. Monitored and evaluated operation of power plant process systems.
Responsible for improvement of systems availability and performance. Performed and planned maintenance activities. Outage coordinator during outages. ;
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APPLICABLE ASSESSMENT EXPERIENCE: f
- Service Water System Operational Performance Assessment (SWSOPA), Surry, 4/94. J
-Instrument Air Safety System Functionallospection (SSFI) Readiness Assessment at i Surry, 1988.
- Auxiliary Feedwater Safety System FunctionalInspection (SSFI) Readiness Assessment at North Anna ,1987 l
- Auxiliary Feedwater Safety System Functional Inspection (SSFI) Readiness Assessment at Surry ,1986 ;
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I STUART M. KLEIN, P.E. j ,
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' CDUCATION B.S., Pennsylvania State Univers'ty,1960 PROFESSIONAL AFFILIATIONS ,
i Rsgistered Professional Engineer, Commonwealth of Pennsylvania ;
SUMMARY
OF QUALIFICATIONS Mr. Klein's diverse experience spans over 31 years of engineering design in such areas as nuclear power plant systems and mechanical equipmer.t design, industrial mschanical design, design review, and project management. He has more than 12 ,
years of nuclear power plant project assignments while employed with a major i crchitect engineering firm. His work has included the detailed design of mechanical i systems with assignments of increasing supervisory and management responsibilities. -l Ha has participated in numerous design inspections (SSFI) with the NRC as well as j utility sponsored SSFI programs. Mr. Klein was responsible for the development of l studies related to configuration management controls at operating nuclear plants for :
tha NRC and continues efforts in these areas to support the nuclear power industry.- l PROFESSIONAL EXPERIENCE j Ogden Environmental and Energy Services Co., Inc.
1984 - present Principal Engineer. Mr. Klein is responsible for overseeing and directing the activities related to mechanical engineering design and desigr. review of power plant process . -
systems. In addition, he has served as a consultant to the Nuclear Regulatory Commission and participated in the design review of numerous safety-related nuclear plant systems, including inspections at FitzPatrick, Calvert Cliffs, and Vermont Yankee (Service Water System Operational Performance inspection), Palisades, Crystal River, D. C. Cook, Calvert Cliffs, Fermi 2, Farley, and Palo Verde (SSFI), and Dresden (SSOMI). Mr. Klein developed a training program for the NRC on vartical-slice inspection techniques used in SSFis which he presented at the NRC Regional Offices and NRC Headquarters. He was involved in the development of studies related to configuration management at operating nuclear plants for the NRC. The results of those efforts have been published by the NRC in NUREG/CR-5147, " Fundamental !
Attributes of a Practical Configuration Management Program for Nuclear Plant Design Control." Mr. Klein has provided support and consultation to the NRC and utilities (Arizona Public Service and Consumers Power Company) in the development of Design Basis Documents. He has served as Design Basis Program Coordinator directing all l 1
Klein - 1
I activities related to the development of Palo Verde Nuclear Generating Station Design Basis Documents for Arizona Public Service. :
Mr. Klein has also written and served as Project Manager to develop Design Basis Documents for several safety related systems at Fermi 2 Nuclear Plant.
Rscently, Mr. Klein participated as lead mechanical reviewer in a mock inspection of i ths emergency service' water system for NYPA at FitzPatrick Nuclear Power Plant.
This inspection was conducted in preparation for the forthcoming NRC Service Water '
Opsrational Performance inspection (SWOPI). He has also participated as the mschanical reviewer in a utility-initiated EDSFI service water system. Mr. Klein also
~
psrformed an evaluation of the Calvert Cliffs Nuclear Power Plant safety-related salt water pumps.
United Engineers & Constructors, Inc., 1972-1984 ;
Supervising Engineer. Mr. Klein had lead responsibility for the Mechanical Group, Site Support Engineering for the Seabrook Nuclear Power Station. Activities included work ,
in sll areas of the plant, both safety related and the balance of plant systems, e.g., .
main steam, circulating water, feedwater systems, and related auxiliary systems. Mr.
Kisin's personal responsibilities included directing the work of the engineers and designers; reviewing and approving drawings, documents, and specifications for plant modifications; and, in general, supporting the construction and start-up efforts to complete the Seabrook project.
During this period, Mr. Klein originated the system designs for the safety-related l station service water system and a number of other cooling systems, e.g., the component cooling water system. He completed extensive trade-off studies to determine optimum system concepts, equipment sizes and parameters for wet and dry '
cooling towers, heat exchangers, pumps, etc. He developed final detail designs and directed procurement activities associated with these systems. Much of the , ;
conceptual work for these activities was described in a paper entitled " Emergency Shutdown Cooling Towers - Considerations in the Evolution of an Optimum Tower .
Design," which was published in Nuclear Safety.
Mr. Klein appeared before the NRC Staff to substantiate the design of essential cooling water systems.
Westinghouse, Bettis Atomic Power Laboratory, 1969-1972 Senior Design Engineer. Mr. Klein was responsible for the design of nuclear reactor plant fluid systems for NIMITZ class nuclear aircraft carriers. He conducted design
. analysis to assure successful hydraulic and thermal performance of the systems. 1 Klein - 2
United Aircraft Corporation, 1963-1969 Design Engineer. Mr. Klein was responsible for the design of aircraft propeller systems and components, pitch change mechanisms, and blade retention systems. He also dnsigned aircraft air inlet control systems, hydraulic actuators, and servomechanisms.
Ha was involved in design tradeoff studies to determine optimum control configurations.
North American Aviation, Inc., 1962-1963
.Research Engineer. Mr. Klein was involved in the design of the engine actuation system for the Saturn 11 Space Vehicle.
United Aircraft Corporation, 1960-1962 Development Test Engineer. Mr. Klein was responsible for the development testing of jet engine fuel control systems and hydro-mechanical feedback control servo-machenisms. He also was involved in the development testing of precision control system components, e.g., flapper control valves, servo controlled linear throttle valves, linkages, pressure control valves, and force balance systems.
PUBLICATIONS NUREG/CR-5147, " Fundamental Attributes of a Practical Configuration Management Program for Nuclear Plant Design Control," U.S. Nuclear Regulatory Commission, 1988.
l Klein - 3 I
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STEPHEN J. JAREMA, JR., P. E.
EDUCATION B. A., Psychology,1964, Washington Square College of Arts & Sciences B. S., Nuclear Engineering,1967, New York University M. S ., Nuclear Engineering,1 g69, New York University Completed all course work for Ph.D. at New York University, 1969-1970 One year at Fordham School of Law, 1963 1964 Four years at Courant Institute of Mathematical Sciences, 1970-1974, N.Y.U.
PROFESSIONAL REGISTRATIONS Pennsylvania, New Jersey
SUMMARY
OF EXPERIENCE Mr. Jarema has over 19 years of experience solving the most difficult of problems confronting nuclear electric power generating stations -- problems affecting inechanical, nuclear, structural, mechanical services and l&C disciplines. Expertise also includes verbal skills, preparation of SARs, etc., testifying at hearings.
PROFESSIONAL EXPERIENCE United Engineers & Constructors Inc.
Mechanical / Nuclear Engineer,1974 Present Public Service Co. of New Hampshire: Seabrook Station. Wrote the Design Basis ,
Document for the Primary Component Cooling Water System. Wrote Subsections 6.2.3, Secondary Containment Functional Design, and 6.2.5, Combustible Gas Control in Containment, for the FSAR Participated in all analyses for these subsections.
Verified the ability of the cooling tower to function within the guidelines of Regulatory Guide 1.27, Ultimate Heat Sink. Assessed the consequences of oil fires in the Diesel Fuel Storage Building or in the containment, fueled by reactor coolant pump lubricant.
Carolina Power & Light Company: Brunswick Nuclear Plant Units 1 and 2. Determined radiological doses in all areas of the plant where post-accident samples are required to be taken for the purpose of shielding design. Participated in a comprehensive i system by system determination of prevailing pressures and temperatures during normal operation, accident and abnormal conditions in order to perform pipe stress analysis to provide input to the design of safety related pipe and equipment supports.
Jarema 1 1
, Power Authority of the State of New York Indian Point 2. Determined radiological doses in all areas of the plant where post-accident samples are required to be taken for the purpose of shielding design.
Philadelphia Electric Company: Peach Bottom 2 and 3. Performed analysis of pressures and flows in all fluid systems to demonstrate compliance with NRC Generic Letter 8910.
Pennsylvania Power & Light: Susquehanna Steam Electric Station. Compiled the design inputs for the replacement of the RWCU pumps with improved sealless models precluding leakage. Performed calculation of heat removal capacity of fan cooler retubed with stainless steelin place of copper-nickel.
Washington Public Power Supply System WNP-1 and 4. Wrote Subsections 6.2.1, Containment Functional Design,6.2.5, Combustible Gas Control in Containment, and 9.2.5, Ultimate Heat Sink, for the FSAR. Performed all analyses required for these subsections.
United States Dept. of Energy: Savannah River Site. Wrote SAR Chapter 7, Radioactive Waste Confinement and Management, for the HB-Line, which processes fissile materials.
Hoeganaes Sponge Iron Corp., Riverside, NJ plant. After visit to client's facility, analyzed cause of hydrogen explosions occurring in annealing furnaces and determined design parameters for implementation of mitigating measures.
COMPUTER CODES WRITTEN ACTNIODE: Determines the equilibrium radionuclide concentration in the vessels and equipment of any interconnected network.
HVAC: Determines the design temperatures for equipment and instrumentation qualitication under any mode of operation of the HVAC system including accident or loss of any part of the total capability.
H2KK: Determines the adequacy of nitrogen injection rates and necessary venting rates / schedules for the combustible as control system of a BWR containment. . c, GWTD: Calculates the time-dependent radio contaminant concentrations in the groundwater following an accidental spill or leak.
FISH 2: Evaluates the impact on river fish populations of pra plant ser,?e water intake.
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'ONVERT:. Determines the coward vertical air velocity developing over a spray -pond
,erving as the ultimate heat sink of a nuclear power plant DESPOND: Calculates the drift loss fraction applicable to a spray pond serving as the ultimate heat sink of a nuclear power plant.
SPOOL: Analyses local thermal effects of densifying spent fuel pool loading.
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JAMES R. ROTH Senior Staff Engineer, Corporate Nuclear Safety EDU C ATION:
B. S., Nuclear Engineering, Pennsylvania State University,1981 LICENSES / SPECIAL TRAINING:
Senior Reactors Operators License, North Anna Power Station,1988 Shift Technical Advisor, North Anna,1983 Root Cause Evaluation,1992 EXPERIENCE:
1981 - 1994 VIRGINIA POWER COMPANY Senior Staff Enaineer. Corocrate Nuclear Safety (4/91-Present). Responsible for preparing Industry Operating Experience analysis reports, independent reviews of safety analyses and for performing assessments.
Shift Technical Advisor. North Anna (10/83-4/91). Responsible for On-Shift STA duties, performance and review of 10 CFR 50.59 evaluations and member of on site ISEG. Other responsibilities included analysis of Industry Operating Experience documents associated with North Anna and in-house Deviation Report corrective action processing and review. Performed post event evaluations (root cause, HPES, reactor trip) as necessary and assisted in performing departmental self assessments.
Encineer. North Anna (6/81-10/83). Responsible for development, review and r>
performance of surveillance tests.
APPLICABLE ASSESSMENT EXPERIENCE:
Safety Evaluation Assessment, Surry and North Anna,11/93
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JOEL 1. KELLY Senior Staff Engineer, System Engineering, Surry Power Station EDUCATION:
B. S., Nuclear Engineering, North Carolina State University,1976 i
LICENSES / SPECIAL TRAINING:
6 l Senior Reactor Operator's License, Surry Power Station,1980.
EXPERIENCE:
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1979 - 1994 VIRGINIA POWER COMPANY, SURRY POWER STATION Senior Staff Engineer. System Enaineerina (1988 - Present). Currently assigned as System Engineer for the Main Turbine and Main Steam systems. Previously responsible for the Liquid and Gaseous Waste systems. Functioned as backup System Engineer for the Service Water system. Responsible for supporting virtually all aspects of plant operation and maintenance relating to assigned systems.
Technical Advisor / Senior Engineer. OA/QC Deoartment. (1986 - 1988). Provided technical support to QA/QC activities and participated in SSFl type inspections.
Senior Enaineer. Safetv Enaineerina Nuclear (1980 - 1986). Duties with the Safety i
Engineering Staff included the investigation and preparation of LERs, Reactor Trip l Reports, and the review and dissemination of industry information from Nuclear Network. Served four years as a Shift Technical Advisor (STA).
Associate Enaineer. Plant Enaineerina Deoartment (1976 - 1979). Prepared Design Modification Packages, participated in post refueling Startup Physics Tests including positions as lead engineer, Reactor Engineer and Performance Engineer. Performed flux mapping and provided assistance in nuclear instrument calibrations. Overall l responsibility of administration of the Periodic Test Program as Performance Engineer.
l APPLICABLE ASSESSMENT EXPERIENCE:
- Safety System FunctionalInspection (SSFI) on the Auxiliary Feedwater system, .
Team leader for SSFI on check valves
THOlVIAS A. KENDZIA, JR.
System Engineer, Corporate Nuclear Safety EDUCATION:
B. S., Nuclear Engineering, Rensselaer Polytechnic Institute,1979 LICENSES / SPECIAL TRAINING:
Shift Technical Advisor, Surry Power Station,1984 Senior Reactor Operator's License. Surry Power Station,1988 Professional Engineer License, Virginia,1984 EXPERIENCE:
1992 - 1994 VIRGINIA POWER COMPANY CORPORATE OFFICES System Encineer. Corocrate Nuclear Safety (8/92-Present). Responsible for performing assessments (See Assessment Experience below), performing independent reviews of Safety Evaluations, performing evaluations of industry operating experience, coordinating Virginia Power Westinghouse Owners Group activities as the alternate to the Primary Representative. Developed program for reviewing NRC Violations at other utilities for applicability to Virginia Power.
1973 1990 SURRY POWER STATION, VIRGINIA POWER COMPANY Suoervisor. Station Nuclear Safety (3/89-8/92). Supervised on shift STAS, HPES Coordinator, and Operating Experience Group. Responsible for evaluation ano I
resolution at the station of industry operating experience items, evaluation and acceptability of resolutions to station problem reports, and the root cause process, implemented revisions of station processes for: problem reporting, industry operating I
experience review, reportability determination, root cause evaluation, and trending for station problem reports. Developed programs for monitoring system / component status while shutdown , Shutdown Critical Parameters, and At Power Critical Parameters l Shift Technical Advisor (3/86 3/89). STA on Operations shift providing Engineering i support for design, system, and ISI. Responsible for reactor trip and event root cause analyses. LER writing, independent review of operability and reportability decisions, I safety evaluation writing. Functioned as Containment Coordinator, Refueling Outage Coordinator, and Refueling SRO. After 7/88,25% of time was spent as unit SRO.
l Enoineer. Station Enaineerina (3/81-3/86;. Performed system and component problem l
resolution, general operations and maintenance support, outage support, engineering evaluations and design changes, ISI and ALARA engineering, and STA duties. Lead Engineer for design changes including Charging Pump Service Water System modification. Qualified as level 2 visual inspector for ISI program. STA on shift rotation until 3/86. Acted as Lead Design Control Engineer for 6 months. Virginia Power lead for Drawing Update Project.
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THOMAS A. KENDZIA, JR.
System Engineer, Corporate Nuclear Safety (cont.)
9/79 - 3/81 NEWPORT NEWS SHIPBUILDING AND DRY DOCK CO.
Enaineer. Nimitz Class Reactor Plant Fluid Svstems Deoartment. Performed engineering evaluations, problem diagnostics, and system evaluations. Qualified as a test engineer and developed initial refueling plan for Nimitz Class Carriers.
APPLICABLE ASSESSMENT EXPERIENCE:
Team Leader:
- Overpower Root Cause Evaluation, Surry,5/94.
- Nuclear Safety Policy implementation, Surry and North Anna,12/93.
-INPO Evaluation Readiness Assessment (Maintenance Lead),6/93.
- Switchgear Fire Event Review, North Anna,11/92.
Participant:
- Service Water System Operational Performance Assessment (SWSOPA), Surry,4/94.
- Procedure Assessment, Surry and North Anna,11/92.
Utility Exchange Assessments:
- INPO Peer Evaluator for Operating Experience, Palo Verde (APS),1991.
GREGORY L. PRESCOTT Assistant Shift Supervisor (SRO), Surry Power Station EDUCATION:
Attended Memphis State University Nuclear Operator Development Program (Center for Nuclear Studies),6/79 - 12/79.
LICENSES / SPECIAL TRAINING:
Senior Ret.ctors Operators License, Surry Power Station,12/88 Reactor Operators License, Surry Power Station,6/82 EXPERIENCE:
1979 - 1994 VIRGINIA POWER COMPANY, SURRY POWER STATION Assistant Shift Suoervisor (3/89 - Present). Supervise daily Operations of a two unit nuclear plant. Assisted in development and implementation of Operation / Maintenance Advisor Program 1/93 - 7/93.
Control Room Ooerator (6/82 - 3/89). Control room operator for 2 unit nuclear plant.
Control Room Ooerator Trainee (12/79 - 6/82). Operated equipment at a 2 unit nuclear plant while in training as a control room operator.
APPLICABLE ASSESSMENT EXPERIENCE:
- INPO Peer Evaluator in Operations at Shearon Harris, 8/91.
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t NEIL S. TURNER Licensed Control Room Operator (RO), Surry Power Station EDUCATION:
Attended Center For Nuclear Studies, Memphis State University,7/82 LICENSES / SPECIAL TRAINING:
Reactor Operator's license,12/88 Root Cause Evaluator Training EXPERIENCE:
1982 - 1994 VIRGINIA POWER COMPANY, SURRY POWER STATION Licensed Control Room Ooerator (1988 - Present). Control room operator for 2 unit nuclear plant, involved in several special projects while in this position which included working on outage scope planning and integration of Service Water maintenance into the outage scope for refueling outages in 1991 and 1992. ' Also involved in construction tie ins for the installation of two new, safety related control room chillers and their Service Water connections. Performed Deviation Report research and response preparation as a member of the Operations Review Board.
Non-licensed Control Room Ooerator (1982 - 1988). Operated equipment at a 2 unit nuclear plant. Assigned as Operations Contact for Construction during the first Unit 1 Circulating Water (CW) and Service Water outage. This required detailed work scope planning and tagout preparation for CW and SW systems, draining the systems, and returning them to service after maintenance.
APPLICABLE ASSESSMENT EXPERIENCE:
Opersoons Department Assessment (w/OA), Surry,1991.
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MANUEL O. BANDEIRA EDUCATION B.S.M.E., Newark College of Engineering,1971 M.S.M.E., New Jersey Institute of Technology,1980
SUMMARY
OF QUALIFICATIONS
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Mr. Bandeira has 20 years of engineering experience in the nuclear and the fossil energy field. His experience encompasses architect-engineer design and engineering, project management, construction, engineering management and _ -
supervision, configuration management, procedure and technical standard development, procurement engineering, INPO ISI and IST evaluator, EPRl/NCIG Steering Committee, and other responsible industry committee memberships.
PROFESSIONAL EXPERIENCE Ogden Environmental and Energy Services Co., Inc.
1991 - present Consultant. Mr. Bandeira's principal duties involve design review and inspection activities for nuclear utility clients. These activities include Safety System Functional inspections (SSFis). He has participated as lead mechanical reviewer in utility sponsored SSFis at Grand Gulf and Perry; the inspections were part of an overall program for the BWR-6 Owners Group. He has also participated in the review of mechanical calculations for the Grand Gulf Nuclear Station's Calculation Assessment. In this effort, Mr. Bandeira's review included many safety-related system calculations.
Mr. Bandeira has also participated in an Environmental Qualification Assessment at Palo Verde Nuclear Generating Station. He also participated in an assessment of the implementation of Generic Letter 89-10 at the Vermont Yankee Nuclear Plant; the assessment focused on MOV performance issues.
Public Service Electric and Gas Company 1971 'i991 Nuclear Engineering Standards Manager. Mr. Bandeira was the department manager responsible for configuration management and control, procedure development and control, programmatic and technical standards development, procurement engineering, and CAED. Mr. Bandeira mcnaged a budget of $7 million and 57 engineering, technical, and administrative personnel.
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Project Manager. Mr. Bandeira was responsible for all engineering, design, construction, and testing phases of the CRD maintenance facility at Hope Creek.
INPO Evaluator. As a nuclear plant technical department evaluator, Mr. Bandeira was responsible for performance monitoring, reactor engineering, engineering administration, surveillance testing, in-service inspection, in-service testing, and design change effectiveness.
l Nuclear Plant Engineering Manager. Mr. Bandeira provided reactive short-term l engineering support for Salem and assisted in the development and implementation l of the systems engineering concept at Salem.
l l Assistant Manager, Nuclear Systems Engineer. Mr. Bandeira was responsible for nuclear and balance of plant systems (e.g. service water systems) engineering management. He participated in the PWR Owners Group and on other industry committees.
Salem Nuclear Systems Group Supervisor. Mr. Bandeira supervised nuclear plant system evaluations, safety analysis, calculations, directives, design analysis and -
development, and operating plant engineering support. He was in direct supervision of Salem Nuclear Systems' HVAC group.
System Sponsor Engineer. Mr. Bandeira was responsible for total sponsorship of l
several nuclear and fossil plant systems from design, component procurement, installation, start-up, and operating technical support. Major projects include spent nuclear fuel storage rack design and installation, and reconstruction of a damaged fossil turbine generator unit, which included all mechanical systems design, installation, and total project control, testing and start-up.
Field Engineer. Mr. Bandeira was responsible for engineering support and interference resolution during construction of the Salem Nuclear Plant.
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C. R. Scllay Test Engineer, Surry Power Station EDUCATION:
l Attended Christopher Newport College,1962 LICENSES / SPECIAL TRAINING:
i SRO training,3/84 3/85 l EXPERIENCE:
1974 1994 VIRGINIA POWER COMPANY, SURRY POWER STATION Test Encineer. Encineerino Testino Grouo (4/85 - Present). In charge of planning and conducting all aspects of Post Modification Testing. Coordinator for Post Maintenance Testing. Responsible for establishing PMT criteria for I & C testing.
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Assistant Instrument Suoervisor (12/79 - 2/84). Supervised daily activities of 10 l instrument technicians.
Instrument Technician (4/74 - 11/79). Fully progressed through I & C development l program. Performed testing and maintenance activities.
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- 1966 1974 U.S. NAVY l
l OA Insoector. Submarine Base New London (7/71 - 3/74). Engineering & Repair department.
Reactor Ooerator. USS Beraall (6/69 - 6/7 J.
Submarine School Trainee. New Londe , Conn. (3/69 5/69)
Navv Nuclear Power School Trainee 1/68 - 2/69) l Electronics Technician. USS Randoloh (7/67 - 12/67).
I Electronics Technician Trainee (7/66 - 6/67).
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1964 - 1966 NEWPORT NEWS SHIPBUILDING AND DRYDOCK CO.
Electrical Hefoer (1964 - 1966).
APPLICABLE ASSESSMENT EXPERIENCE:
- Post Modification Testing Assessment at North Anna,1991.
INPO Peer Evaluator on Post Modification Testing at Oconee Nuclear Station,1993.
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KENNETH C. PIER Senior Quality Specialist, Quality Assurance EDUCATION:
General Equivalency Diploma,1978 Tidewater Community College J. S. Reynolds Community College LICENSES / SPECIAL TRAINING:
General inspector Level 11 (Certification current)
Visual Weld inspector Level ll (Certification current)
Auditor (Certification current)
Nuclear Power Operations Development Program,11/82 Reactor Operator licensing training, 4/84-4/85 Nuclear Power Mechanical Development Program,12/88 EXPERIENCE:
1979 - 1994 VIRGINIA POWER COMPANY, SURRY POWER STATION Senior Qualitv Soecialist. Quality Assurance (5/89 Present). Participated in station audits, performance based assessments, inspections of maintenance, design change and/or operations activities.
Senior Quality Control Insoector (19851989) Conducted performance based assessments and OC inspections. Performed quality inspections for construction, station and vendor activities to verify compliance to codes, standards, and procedures.
Assisted in development of OA Department procedures and instructions. Reviewed Work Orders and Engineering Work Requests, validated Design Change packages and Engineering Work Requests.
Ooerator (19791985). Routine and special operations of Surry 1 and 2. Routine responsibilities in plant parameter documentation, Periodic Test performance, and plant equipment operation. Performed hydro's and pressure / flow test of valves components & piping of various new and pre-existing systems.
1974 - 1979 OTHER EXPERIENCE Ooerator. Temoerature Control (1978-1979). Smithfield Packing. Operation and maintenance of refrigeration equipment and plant auxiliary boilers.
Shoofitter (1376-1978). Newport News Shipbuilding and Drydock Corporation.
Certified tack welder, gas torch and carbon arc. Interpreted blue prints in work on Naval and commercial vessels.
Ooerator. Temoerature Control (1974-1976). Gwaltney's Meat Packing. Operation and maintenance of refrigeration equipment and plant auxiliary boilers.
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KENNETH C. PIER Senior Quality Specialist, Quality Assurance (cont.)
APPLICABLE ASSESSMENT EXPERIENCE:
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- Service Water System Operational Performance Assessment (SWSOPA) Response Team, Surry, 4/94.
- Participated in 24 assessments since 1990,7 in Operations,6 in Engineering,6 in Maintenance l
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l DAVID WOOTTEN I
Senior Staff Engineer, Corporate Nuclear Safety EDUCATION:
B. S., Mechanical Engineering, Maritime College of Fort Schuyler,1975 Nuclear Engineering Course Work (22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />), Memphis State,1982 LICENSES / SPECIAL TRAINING:
Shift Technical Advisor, Palo Verde,1981 Senior Reactors Operators License, Palo Verde,1984 Root Cause Analysis,1985 INPO Technical Support Evaluator Qualification,1986 Engineer in-Training (EIT) Test for PE,1991 EXPERIENCE:
1989 - 1994 VIRGINIA POWER COMPANY Senior Staff Engineer. Corocrate Nuclear Safety (6/91-Present). Responsible for performing assessments, (See Assessment Experience below), preparing Industry Operating Experience analysis reports and independent reviews of safety analyses.
Senior Consultant to Virainia Power CNS C eoartment (9/89-6/91). Assisted in upgrading the Operating Experience Program. Setup and trained personnel on the use of Industry databases (Tech Library, NPRDS, NRC Index, LER, Listen / Trends, I etc.). Performed self assessments of specialized areas (e.g. In house Deviation l Reports, Generic Letters,10 CFR Part 21, etc.). Responsible for the analysis of Industry Operating Experience documents associated with North Anna and Surry Power Stations.
1980-1989 ARIZONA PUBLIC SERVICE COMPANY Lead Encineer (4/89 9/89). Responsible for the coordination of five Senior Engineers and the industry Operating Experience Program, including the SOER Reverification Project. Responsibilities included upgrading the Industry Operating Experience Program.
Senior Enaineer (8/87-4/89). Responsible for developing the department mission statement position descriptions, and budget for a new department (Technical Data).
Upgraded the Performance Monitoring, Management Observation, Disciplinary Action, Fitness for Duty and Chemical Control Programs.
Project Manager Assigned to INPO (10/85-8/87). Assisted in development and implementation of the Human Performance Evaluation System (HPES). Developed and analyzed the HPES database. Spearheaded the events and causal factors chart used for the Chernobyl SOER.
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DAVID WOOTTEN Senior Staff Engineer, Corporate Nuclear Safety (cont.) .
Actino ISEG Suoervisor (8/84-10/85). Developed, created and implemented programs to ensure ISEG performed routine surveillances, trending of in-house events and -
performance of specialindependent investigations.
Shift Technical Advisor. Palo Verde (5/81-8/84). Member of the Safety Evaluation Task Force for all TMI related issues. Provided critical reviews of the Standard Technical Specifications do determine applicability to Palo Verde and performed 50.59 reviews.
Designed the Unit 1 NSSS Flush Water Hold Tank and delivery System.
1975 - 1981 OTHER EXPERIENCE Ooeratino Betterment Encineer (4/80-5/81). Four Corners Plant. Prepared plans, designs, specifications, cost estimates and system descriptions.
Senior Field Engineer (2/77 4/80). Schlumbereger Well Services, Farmington, NM.
Responsible for logging outfit, supervision of crew members and client relations.
First Assistant Engineer (2/76-8/76). Seal Craft Operators, Inc., Galveston, TX.
Responsible for all engine room activities on a diesel geophysical research vessel. -
Third Assistant Marine Engineer (8/75 2/76). Military Sea Lift Command, Oakland, Ca.
APPLICABLE ASSESSMENT EXPERIENCE:
Team Leader:
- Charging pump operability event review, North Anna,6/93 I & C Performance Review, North Anna,4/93
- INPO evaluation readiness (Engineering Lead) , Surry and North Anna,1992 l & 1993 l - Switchyard activities, Surry and North Anna,4/92 Participant:
- Service Water System Operational Performance Assessment (SWSOPA),
Surry, 4/94.
- RCE Program Assessment, Surry and North Anna,12/93
- Maintenance Audit, Surry and North Anna,10/92.
Electrical Distribution System Functional Assessment, Surry,5/92.
i NUMARC Shutdown issues, Surry and North Anna,1/92.
Component mispositions/ transpositions, Surry and North Anna,10/91 INPO Technical Support evaluations, eight plants, 10/85-8/87.
INPO Technical Exchange Visit with Electricit6 de France, Paris,10/86.
Utility Exchange Assessments:
Engineering assessment, Diablo Canyon (PG&E),6/93.
Engineering assessment, Corporate (CP&L),7/92.
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i ATTACHMENT 2 :
CORRECTIVE ACTIONS SCHEDULE l
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12/12/94
(]) North Anna Power Station Commitment Trackine System Action Report
'N J VIRZIN A FOWEn Listing of Commitments for CTS Items that are Open Due Date Denartment Source / Description of Action to be Performed Status Number 08/18/95 STAT 10N Service Water System Operational Performance Assessment OPEN 02-94 2229 001 Unit 1 ENGINEERING gg.7g 1) include 1-SW-ll39/1067/1070 to the IST Program and cycle every refueling,2) Develop valve operability test,3) Develop irr to perform each cold shutdown,4) Request CME calc. (re: flow rate for ppg in the event of a non-seismic CA ppg failure, and 5) Revise testing ramts IAW revised 06/30/95 STAllON Service Water System Operational Performance Assessment OPEN 02 94-2229-002 Unit l&2 ENGINEERING ggy Evaluate the design function of the SW pump screens by either: 1) downgrading them to non-safety,2) removing them, or 3) requiring them to be operrble.
06/30/96 FIAT 10N Service Water System Operational Performance Assessment OPEN 02-94-2229-003 Unit I&2 ENGINEERING ggy Revise the following SW DBD open items: 1) SDBD Table 6.1-4 to correctly state SW Gow values for CC flow to ilXs on non-accident unit,2) Basis for ASW pump alarm setpoint,3) Setpoint for SW return header flow and 4) flasis for ASW hich pressure setroint.
11/30/95 STAT 10N Service Water System Operational Performance Assessment OPEN 02-94-2229 004 Unit l&2 ENGINEERING ggy Obtain manual pipe wall measurements to address minimum wall thickness of old SW return lines.
Re-evaluate installing blind flanges on these lines.
01/30/95 STA110N Service Water System Operational Performance Assessment OPEN 02-94 2229-005 i
Unit 1&2 ENGINTHING '
ggy Evaluate the need to revise 0-l'T-75.14 to add monitoring points.
10/28/94 OUTAGE Service Water System Operational Performance Assessment CLSD 02 94-2229-006 Unit 1&2 PLANNING ggy Schedule performance of PMs for disassembly, inspection and refirbushment of the main / auxiliary SW pumps as per RCM Program. ;
12/15/94 PROCEDURES Service Water System Operational Performance Assessment CLSD 02-94-2229-007 Unit I&2 ggg Evaluate development of a new Corrosion Coupon Procedure / Program to replace SW Corrosion monitoring coupons.
12/15/94 STAT 10N Service Water System Operational Performance Assessment OPEN 02-94 2229-008 Unit 1&2 ENGINEERING Revise IST Program Basis " Safety Function" for RSilX check valves so that it reflects that the SITM valves open to proyude flowpaths for cooling water to the RSilX. They have no safety function to close.
08/18/95 STAT 10N Service Water System Operational Performance Assessment OITN 02-94 2229-009 Unit 1&2 ENGINEERING Initiate a revision to UFSAR Section 9.2.1.3.2 to delete reference to isolating a ruptured SW ggg header with the header check valves.
08/30/95 STA110N Service Water System Operational Performance Assessment OPEN 02-94-2229-010 Unit 1&2 ENGINTIRING Verify accuracy f the ruled marker m unted n the side of the SW pump house and develop a plan SITM to ensure it can be read below normal water level.
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( ) North Anna Power Station Commitment Trackine System Action Report vccisiA POWEg Listing of Commitments for CTS Items that are Closed Due Date DImartment Smitee/ Description of Action to be Performed Status Number 11/18/94 OIHAT10NS Service Water System Operational Performance Assessment CLSD 02 94-2229-011 Unit l&2 Upd te VPSLL Breaker Load List to include I/2-EP-CB-Il7.
SITM 12/16/94 OPERA 110NS Service Water System Operational Performance Assessment OPEN 02-94-2229-012 Unit 1&2 Correct various labeling discrepancies: 11715-FM-72il, Sht. I, CLDR#72094,11715 FM 78A, SITM Sht.1. CLDR#, ll715-FM 78A, Sht. 2, CLDR#72594, DCR 94-642, il715 FM 78A, Sht. 3, CLDR#72094 I1715-FB-040D. Sht. 2. CLDR#s72894 and 72994.
11/18/94 OPERAT10NS Service Water System Operational Performance Assessment CLSD 02-94 2229 013 Unit 1&2 Correct various labeling discrepancies: 1) i1715-FB 040D, Sht.1,Rev 23, CLDR 72794,2)
SITM missing labels on I-SW-277 &331, CLDR# 72794 and 72894, and 3) labeling of Cil gear box coolers per 11715-FM-780. rht 1 & 2. CLDR# 72694 (DCR 92-7273).
12/16/94 OPERAT10NS Service Water System Operational Performance Assessment OPEN 02-94-2229-014 Unit 1&2 Correct various labeling discrepancies: 1) Il715 FM 78B, Sht. 3,2) abandoned label on drain SITM valves. CLDR #s 73094 thru 73394, and 3) evaluate use of control room " stick on" markings that are outside the Operator Aid Program.
11/04/94 PROCEDURES Service Water System Operational Performance Assessment CLSD 02-94-2229 015 Unit i Rnise 1400-4A to incorporate SW pump discharge pressure.
SITM 12/09/94 YTA110N Service Water System Operational Performance Assessment CLSD 02-94-2229-016 Unit i ENGINEERING Evaluate decision to disconnect the I-SW.RM -108 DR trend "SW to take radiation monitor" power SITM supply connector. (REA 93 289) 10/31/94 STA110N Service Water System Operational Performance Assessment CLSD 02 94-2229-017 Unit i ENGINEERING Evaluate the need for a irr to test iX-ISWN08, since its failure could challenge an entire train of SITM safety systems.
12/15/95 STAT 10N Service Water System Operational Performance Assessment OPEN 02-94-2229-018 Unit I&2 ENGINEERING Evaluate whether calc. ME-062 Rev. O dated 8/12/85 (addenda thru 5/1/87),1) contains the most SITM conservative design inputs for FRTW and 2) whether DTDRYO/ Pill considers meteorology that will result in minimum spray enoling per RG 1.27. Rev. 2. In6.
12/15/95 STAT 10N Service Water System Operational Performance Assessment OPEN 02-94-2229-019 Unit 1&2 ENGINEERING Evaluate SW reservoir minimum spray efficiency (variable CMIN may have impact on calculated SITM spray efficiencies which may be less than 209).
03/30/95 STA110N Service Water System Operational Performance Assessment OIEN 02-94-2229-020 i Unit l&2 INGINEERING f Revise VPAP-0811 to: 1) limit scope of Station Records SW files required to ensure GL 89-13 l SITM compliance, and 2) require appropriate as found conditions to be recorded by photographic means and stored for trending / inspection / cleaning frequency purpo',es.
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\ j North Anna Power Station Commitment Trackine System Action Report FC/N7A PowEn Listing of Commitments for CTS Items that are Open l
Due Dak Department Source / Description of Action to be Performed S. talus Number 12/15/95 TRAINING Service Water System Operational Performance Assessment Olm 02-94-2229-021 Unit 1&2 gg.g.g Evaluate developing special GL 89-13 inspection training for personnel performing inspections and components covered by GL 89-13. Topics should include GL 89-13 requirements, commitments. recognition of asiatic clams /rebra mussels, and expectations for documentmg '
01/13/95 PROCEDURES Rrvice Water System Operational Performance Assessment OPEN 02 94-2229-022 Unit I&2 Develop a PAR to reflect that 2-SW 278 and 2-SW 1027 are normally closed. Develop Pars to SITM i related procedures that rely on these valve being open. ,
11/18/94 OPERAT10NS Service Water System Operational Performance Assessment CLSD 02-94-2229-023 Unit 1&2 issue a DCR to reflect that 2-SW 278 and 2-SW 1027 are normally closed. Place Polymer Addition SITM System in Abnormal Status until the DCR is complete.
06/30/96 STAT 10N Service Water System Operational Performance Assessment OiG 02 94-2229 024 Unit I&2 ENGINEERING Implement IR # 7006,(GL 89-13) Pipe coating Inspection Program to inspect SW SITM piping / component protective coatings.
03/30/95 STA'llON Service Water System Operational Performance Assessment OPEN 02 94 2229-025 Unit I&2 ENGINEERING issue a task to Calgon to evaluate compatibility of SW chemicals with expansion joint materials.
SITM 12/15/95 STA'llON Service Water System Operational Performance Assessment OPEN 02-94-2229-026 Unit 1&2 ENGINEERING Eval. whether Calc ME-062 incorporates: 1) 4.5% upgrading of the core power from 2775 MW SITM
(+12 MW RCP power for 2905 MW) per DCP 86-021. and 2) containment load from LOCA unit having operated at 102% power and all approp. Aux. loads from both units are consistent with RG 12/20/95 STAT 10N Service Water System Operational Performance Assessment OIM 02-94-2229-027 Unit 1&2 ENGINEERING Evaluate functional testing / delta.P testing of MOVs to discharge tunnel.
SITM 10/31/94 STAT 10N Service Water System Operational Performance Assessment CLSD 02-94-2229-028 Unit I&2 ENGINIERING Revise EDS to identify the SR status of CCllX radiation monitoring piping.
SITM 06/30/95 STAT 10N Service Water System Operational Performance Assessment Olm 02-94-2229-029 Unit l&2 ENGINEERING gg.g.g Evaluate cales related to MOV program acceptance criteria.
11/23/94 STAT 10N Service Water System Operational Performance Assessment CLSD 02-94-2229-030 Unit I&2 ENGtMERING Evalu te radiati n m nitor effects on expansion joint qualification and other equipment located in SITM the QSPil basement to withstand potential steam from IIELD of the steam supply to the AFW pump turbine and radiation from SI piping.
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vinciNIA)' POWEn NorthListing Anna Power Station of Commitments for CTS Items Commitment that are Open Trackine System Action R llue Datt Draatiment Snurce/ Description of Action to be PerformcL Statu.s Numb.cr I 1
02/06/95 STAT 10N Service Water System Operational Performance Assessment OPEN 02 94 2229-031 j Unit l&2 ENGINEERING gggg Resolve apparent discrepancy between EQ zone source terms and the source term used for the shielding calc. for the RSilX radiation monitors (Calc estimate on the background radiation in the rad monitor cubes did not confirm background levels were low enough to achieve the minimum i
10/31/95 STAllON Service Water System Operational Performance Assessment OftN 02 94-2229-032 Unit I&2 ENGINEERING gggy Evaluate for impact: 1) cumulative effects of losses thru non-safety related, un isolated pipingon SWS flows,2) basis for the SW pump composite curve used and the methodology used to develop l the curve should be documented. and 3) potential degradation. ;
10/31/95 STA110N Service Water System Operational Performance Assessment OIYN 02-94 2229-033 ;
Unit I&2 ENGINIIRING Evaluate for impact: 1) effects of max. SWS pump flow and maximum reservoir temperature on SIThi NPSil in the " Strong" pump /" Weak" pump scenarios, and 2) evaluate / document analysis that sufficient now is available to meet the 4500 rpm design flow to the RSilXs and other safety 12/15/95 PROCEDURES Service Water System Operational Performance Assessment OIYN 02-94-2229 034 Unit I&2 gggg Revise SW chemical addition line break inventory loss procedures to require monitoring of the chemical addition system non-seismic piping added by DCP 85-48-3, as required by the safety ,
analysis. !
12/20/95 STA110N Service Water System Operational Performance Assessment OlYN 02 94 2229 035 Unit I&2 ENGINEERING Revise irr-75.4 to include clear acceptance criteria for the inspection of the screens for SIThi ,
crosion/ corrosion.
I1/15/94 PROCEDURES Service Water System Operational Performance Assessment CLSD 02-94-2229-036 Unit l&2 Revise PM M 10-TS/SA-1 to note concerns identined during performance of the PM.
SITh1 12/09/94 STA110N Service Water System Operational Performance Assessment CIED 02 94 2229 037 Unit 1 INGINEERING gggg Evaluate the need to periodically verify that 1-SW RM 107 and 108 both sample from the SW discharge headers simultaneously since blockage from biofouling, debris or a check valve failure could prevent a sample from being obtained.
12/15/95 PROCEDURES Service Water System Operational Performance Assessment OPEN 02-94 2229-038 Unit l&2 gggg Evaluate revising EOPs to provide guidance for isolating SW to the RSilXs when recirculation spray is secured if credit is taken for isolating SW to the RSilXs during the course of a DilA.
10/31/95 STA110N Service Water System Operational Performance Assessment OIYN 02-94-2229-039 Unit 1&2 ENGINERING Evaluate: 1) EDG frequency history to determine the amount by w hich the acceptable frequency SITh1 band can be narrowed without significantly increasing the predicted EDG failure rate, and 2) whether setpoints for manually adjusting EDG frequency in the APs should be set within the test 06/30/96 STA110N Service Water System Operational Performance Assessment OPIN 02 94 2229 040 Unit l&2 ENGINEERING gggg Evaluate operation of the Cathodic Protection System and the effect of new protective roatings on SW piping cathodic protection design prior to implementing PMs E-II-MISC /M1 and sal.
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(, ) North Anna Power Station Commitment Trackine System Action Report
'O VCGINIA POWEn Listing of Commitments for CTS Items that are Open Due Datt Draarjment Smtree/Descrintion of Action to be Performed Slatus Number 01/13/95 IROCEDURES Service Water System Operational Performance Assessment Olm 02-94-2229 041 Unit l&2 Evaluate revision of 0-AP 27 to require performance of 0-OP-49.6 when SW is placed in service or SITM removed from service for the SFP. Also directions for restoring systems to normal after the need for SW flow to the coolers is eliminated.
03/30/95 STATION Service Water System Operational Performance Assessment Olm 02-94 2229 042 Unit 1&2 ENGINEERING Revise SW pump head reference curve to reflect CPU data.
SITM 01/13/95 IROCEDURES Service Water System Operational Performance Assessment OPIN 02 94 2229 043 Unit I&2 Evaluate revision of AP 39.2 to identify likehhood of water originating from Turbine 13uilding SITM piping tunnel due to a pipe rupture in the tunnel or associated valve pit, which includes an out-of-the-way imrtion of the SW system 02/20/95 STADON Service Water System Operational Performance Assessment Olm 02 94 2229-044 Unit I&2 ENGINEERING Evaluate keeping the SW to AFW crosstic valves open considering the impact (Single failure SITM vulnerability) on RSliX isolation.
09/29/95 IHOCEDURES Service Water System Operational Performance Assessment Olm 02-94 2229 045 Unit 1&2 Revise the following OP-49 series procedures: 1) 0-OP-49.6, Control of CCilX dP. Add P&L re:
SITM 25 psi delta-P. 2) OP-49.2,3) OP-49.3 and 4) OP-49.4.
08/31/95 STADON Service Water System Operational Performance Assessment OPEN 02 94-2229-046 Unit I&2 ENGINEERING Evaluate implementation of TEI recommendations for flushing sump pumps and floor drains.
SITM Ensure that flooding does not adversely impact safety related equipment in the SWVil.
11/30/95 IROCEDURES Senice Water System Operational Performance Assessment OPEN 02-94 2229-047 Unit 1&2 Revise: 1) LOGS t ensure that 1-SW P-lil is consistent with other pumps and 2) operating SITM procedures to reference correct SW pump discharge pressures.
09/29/95 STATION Service Water System Operational Performance Assessment OPEN 02-94-2229-048 Unit I&2 ENGINIHlNG Revise cale. ME-0026 to fully address acceptability of the system design (i.e., components in the SITM SW flowpath are included in the single failure review w hen it is the dcsigned back-up cooling source).
02/01/95 MAIVIENANCE Service Water System Operational Performance Assessment Oim 02-94 2229-049 Unit 1&2 Evaluate preventative maintenance of CC/SW interface valves.
SITM 08/31/95 IHOCEDURES Service Water System Operational Performance Assessment CLSD 02 94-2229 050 Unit I&2 Implement RCM Task #s 37 for instrument air compressor heat exchangers.
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( brth Anna Power Station Commitment Trackine System Action Report Listing of Commitments for CTS Items that are Open VIRZIN/I' POWER Due Date Department Snurst/ Description of Action to be Performed _ Status Number 02/28/95 PROCEDURES Servi a Water System Operational Performance Assessment OPEN 02 94-2229-051 Unit I&2 Evaluatt need for additional OP-21.1 controls when SW is aligned to the Containment Air SITM Recircui . tion Coolers and SW headers.
01/13/95 PROCEDURES Service Water System Operational Performance Assessment OPEN 02-94-2229-052 Umt I&2 gg.g.g Revise AP 5.1. Attachment 15 to reference VPAP-2103 for offsite dose calculations.
03/10/95 IHOCEDURES Service Water System Operational Performance Assessment OIM 02-94-2229-053 Unit I&2 Evaluate incorporating P&Ls into 1/2-AP-5 regarding the potential for sudden increase in radiation SITh1 levels in the Quench Spray Pump ilouse basement.
09/29/95 STA110N Service Water System Operational Performance Assessment OPEN 02 94 2229 054 Unit I&2 ENGIFEERING Evaluate: 1) effectiveness of 1/2-MOP-49.31 to ensure mud / silt have not accumulated on the heat
'MTM exchanger surfaces of the RSilX tutes and 2) determine inspection frequency based on '95 Inspection results.
06/30/95 STAT'ON Service Water System Operational Performance Assessment Ol m 02 94-2229-055 Unit I&2 ENGINEERING Evaluate base CDA scenario and re-run KYPIPE model and revise cale. ME 0317,if worst case SITM scenario is not assumed, 10/31/94 STATION Service Water System Operationa Performance Assessment CLSD 02-94-2229-056 Unit 1 ENGINEERING Implement DUR 94-649 to revise 1-SW-fT-110 from #3 and #4 header bypass flow to correct SITM numbering for flow instrumentation.
07/31/95 STATION Service Water System Operational Performance Assessment Ol m 02-94-2229-057 Unit 1&2 ENGINEERING Revise 0 PT-75.15 to incorporate commitments to inspect SW lines to the spent fuel coolen made SITM in the latest NRC response to GL 89-13.
11/30/95 PROCEDURES Service Water System Operational Performance Assessment Ol m 02-94-2229-058 Unit l&2 Revise AR lE-F5 to address flooding.
SITM 12/15/95 STAT 10N Service Water System Operational Performance Assessment OPEN 02-94-2229-059 Unit 1&2 ENGINEERING Evaluate whether the proposed addendum to Cale. ME-0420 adequately takes into account that SITM maximum calculated reservoir temperatures may exceed 100F when the considerations related to ME-062 (e p., charging pumps and chiller minimum design flow requirements are addressed.
I1/30/95 IHOCEDURES Service Water System Operational Performance Assessment Ol m 02 94-2229-060 Unit 1&2 Evaluate revisions to the following ARPs: ARP IK-G5 for loss of air at SWPil, ARP IJ.B4, ARP SITM IJ E5, ARP IJ-D4 and ll E3, ARP IJ.D3, ARP lj.H3, ARP IE-C7, ARP IK-F1, ARP IK-D4, add P&l re: entering high rad rones. ARP lj.ll6 ARP 11 D5, develop ARP for trubine building valve w 12/12/94
( North Anna Power Station Commitment Trackine System Action Report VlZINII POWEg Listing of Commitments for CTS Items that are Open Due Date RfDarlmtnt Source / Description of Action to be Performed Status Number 08/29/95 PROCEDURES Service Water System Operational Performance Assessment OPEN 02-94 2229-061 Unit 1&2 Revise Chemical Addition System procedures to require appropriate inventory loss monitoring, as SITM required by the safety analysis associated with DCP 85-48-3.
10/31/94 STA*I1ON Service Water System Operational Performance Assessment CLSD 02-94 2229-062 Unit 1&2 ENGINERING Revise CDRR to DCP 92-123 to delete reference to the auto strainers.
SITM 01/12/95 STATION Service Water System Operational Performance Assessment MGMT 02-94-2229-063 Unit 1&2 ENGINEERING Perf rm final Technical Review of DCP 85-48.
SITM 08/29/95 PROQDURES Service Water System Operational Performance Assessment OPEN 02 94 2229-064 Unit 1&2 Evaluate whether EOPs contain guidance on flow to non-normal loads: 1) isolating SW to the SITM RSilXs when RS is secured 2) flow to Containment Air Recirculation Cooling on non accident unit, & 3) documenting analysis that determines flows supplied to SR components when these 02/26/96 STA110N Service Water System Operational Performance Assessment OPEN 02-94-2229-065 Unit I&2 ENGINEERING Evaluate: 1) inspection and cleaning of piping exposed to SW prior to reinstallation of RSIIX SITM relief valves and 2) actions needed to to assure RSilX relief valves will perform as required if they fail to meet established criteria when tested.
10/31/94 STA110N Service Water System Operational Performance Assessment CISD 02 94 2229 066 Unit 1&2 ENGINEERING Evaluate methods to process outstanding DCR in a timely manner (i.e., DCRs 94-1008,93-769, SITM 91-011, 93 770,93-821, 941007, 94-1006, 94-1019,93-743, 93-747, and 94-1020 12/15/95 STA110N Service Water System Operational Performance Assessment OPEN 02-94-2229-067 Unit l&2 ENGINEERING Evaluate whether sufficient conservatism exists in ME-%2, Reservoir Performance Analysis in SITM view of issues / questions raised.
01/31/95 OUTAGE Service Water System Operational Performance Assessment OPEN 02 94 2229 068 Unit 1&2 11ANNING Implement RCM Task #s 38 for instrument air compressor heat exchangers.
SITM STA110N Service Water System Operational Performance Assessment Ol1N 02 94-2229-069 Unit 1&2 ENGINEERING Evaluate methods to process outstanding Configuration Management DCRs in a timely SITM manner (DCRs93-743, 93-747, 93 769,93-770, 93 821).
10/31/95 STA110N Service Water System Operational Performance Assessment OPEN 02 94 2229 070 Unit 1 ENGINEERING Ensure a placard 4 iabricated and installed by the detector cat les for 1-SW-RM 108 that states SITM " Caution Sensitive Equipment"
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! ) North Anna Power Station Commitment Trackine System Action Report
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' VCC/N A POWEg Listing of Commitments for CTS Items that are Open Due Date Denartment Source /Descrlotion of Action to be Performed Status Number 10/31/95 STA110N Service Water System Operational Performance Assessment OIEN 02-94-2229 071 Unit i ENGINEERING gg Develop a IT to verify flow through each of the header lines to I-SW-RM 107 and 1 SW RM-108.
Total Count for Open = 71 l
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