ML20239A104
| ML20239A104 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 08/27/1987 |
| From: | Croley B, Humenansky D, Prince K SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20238F564 | List: |
| References | |
| NUDOCS 8709170079 | |
| Download: ML20239A104 (30) | |
Text
EXPANDED AUGMENTED SYSTEM REVIEW AND TEST PROGRAM (EXPANDED ASRTP)
EVALUATION OF THE SEAL INJECTION AND MAKEUP SYSTEM SUBMITTED BY:
d
~M >l DATE: CS-Lle-87 Kt/TH' PRINCE TEAM LEADER CONCURRENCE:
tM -
1-u4 DATE:
@~lb D
/ DAVID HUMENAN' SKY j
EXPANDED ASRTP PROGRAM MANAGER CONCURRENCE:
DATE: b'
?
Il
/ BOB CROLEY
/
DIRECTOR, NUCLEAR TECHNICAL SERVICES g91{0h p
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TABLE OF CONTENTS Paae Number
1.0 INTRODUCTION
3 2.0 PURPOSE 4
3.0 SCOPE 5
4.0 OVERALL RESULTS AND CONCLUSIONS 6
5.0 SPECIFIC CONCERNS 8
5.1 Acknowledged (Valid) Concerns 8
5.2 Open (Potential) Concerns 8
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6.0 ATTACHMENTS 9
6.1 List of Documei t3 Reviewed 11 6.2 Status of RIs 15 1
6.3 Detailed Observations - Requests for Information 16 i
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EXPANDED AUGMENTED SYSTEM REVIEW AND TEST PROGRAM EVALUATION OF THE SEAL INJECTION AND MAKEUP (SIM) SYSTEM
1.0 INTRODUCTION
l The Rancho Seco Expanded Augmented System Review and Test Program
[ASRTP) evaluation effort involves an assessment of the effectiveness of the System Review and Test Program [SRTP] and an analysis of the adequacy o' ongoing programs to ensure that systems will continue to function properly after restart.
The Expanded ASRTP is a detailed system by system review of the SRTP as I
implemented on 33 selected systems and an indepth review of the engineering, niodification, maintenance, operations, surveillance, inservice testing, and quality programs.
It also conducts a review, i
on a sampling basis, of many of the numerous ongoing verification and review programs at Rancho Seco.
Six multi-disciplined teams composed of knowledgeable and experienced personnel are tasked with performing the Expanded ASRTP.
Each multi-disciplined team consists of dedicated personnel witn appropriate backgrounds to evaluate the operations, maintenance, engineering, and design functional areas.
Independence, perspective, and.ndustry standards provided by team members with consultants, architect engineer and vendor backgrounds are joined with the specific plant knowledge of SMUD team members.
Each team performs an evaluation on a selected system using the same fundamental evaluation techniques employed by the NRC in the ASRTP 1
inspection.
System Status Reports are used as the primary source of leads for the teams.
They are augmented with references to available source and design bases documents as needed.
Team synergism and communication is emphasized during the process in l
l order to enhance the evaluation.
Each team prepares a report for each completed selected system evaluated.
This report is for the Seal Injection and Makeup (SIM) system.
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_ _ _- _ _ _ ___ __ A
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2.0 PURPOSE The objectives of the Expanded ASRTP evaluation are to (1) assess the adequacy of activities and systems in support of restart and (2) evaluate the effectiveness of established programs for ensuring safety during plant operation after restart.
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3.0 SCOPE To accomplish the first objective, the Reactor Plant System team evaluated the SIM system to determine whether:
1.
The system was capable of performing the safety
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functions required by its design bases.
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2.
Testing was adequate to demonstrate that the system would perform all of the safety functions required.
3.
System maintenance (with emphasis on pumps and valves) was adequate to ensure system operability under postulated accident conditions.
4.
Operator and maintenance technician training was adequate to ensure proper operations and maintenance of the system.
5.
Human factors relative to the system and the system'_s supporting procedures were adequace to ensure proper system operations under normal and accident conditions.
To accomplish the second objective, the team reviewed the programs as implemented for the system in the following functional areas:
1.
Systems Design and Change Control l
2.
Maintenance 3.
Operations and Training 4.
Surveillance and Inservice Testing I
5.
Quality Assurance 6.
Engineering Programs The team reviewed a number of documents in preparation for and during the Expanded ASRTP evaluation.
This list of documents is found in Attachment 6.1.
The primary source of leads for the team were the problems identified in the SIM System Status Report.
Various source documents such as the USAR and Technical Specifications and available design bases documents were reviewed as needed to augment the information needed by the team.
The evaluation of the SIM system included a review of pertinent portions of support systems that must be functional in order for the SIM system to meet its design objectives.
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4.0 OVERALL RESULTS AND CONCLUSIONS The more significant issues identified pertaining to the adequacy of the SRTP and the effectiveness of programs to ensure continued safe operations after restart are summarized below.
The summary focuses on the weaknesses identified during the evaluation..3 provides detailed findings by providing the Request for Information (RI) forms that are used by the Expanded ASRTP teams to identify potential concerns during the evaluation.
Section 5.0 lists the specific concerns identified by the teams.
The numbers in becckets after each individual summary or concern refer to the corresponding RIs in Attachment 6.3.
4.1 Potentially, of most concern are generic problems with the effort to refurbish all Motor Operated (MO) Valves.
These problems are not unique to SIM but may affect any system that contains M0 Valves.
The concerns include procedures, testing, engineering, qualified parts and documentation.
No governing or controlling procedure is in place to tie all aspects i
I of the program together.
There are some maintenance procedures that provide instructions for mechanical or electrical assembly /
disassembly but do not document how calculations are to be performed, or how approval and signoff is to be done after the work is completed.
Instructions are not provided to set external limit switches nor functionally test the settings prior to startup.
Settings on valves that interact with other components will not be tested with respect to the interaction with the system prior to startup; however, the valves have been turned over as work completed.
There is no independent verification of calculations or results of work ps.rformed by Nuclear Engineering (NE).
NE is represented by contract personnel on the MO Valve Refurbishment Team.
The NE person responsible for meeting schedule and budgeting, assigning i
work, and resolving problems is the same person that reviews and i
approves calculations and approves work performance.
A check and balance program is not present.
Review of torque value calculations indicated: 1) use of a non-approved computer program, 2) calculations were not signed nor dated, and 3) changes were made that were not initialed and dated.
Basis for the calculations was not documented.
[RI-il9] [RI-121]
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OVERALL RESULTS AND CONCLUSIONS (Continued) 4.2 Programmatic concerns were found with Material Control and Vendor Manual Control during evaluation of the DHR System.
Additional findings during SIM System evaluations strengthen these concerns.
Commercial grade items are being purchased and used on safety-related components without a formal Commercial Dedication Program in place.
The team is concerned that use of commercial parts without proper instructions from an inplace program could jeopardize the environmental qualification of some components.
A sound program should address procurement, receipt, inspection, training, storage, etc.
The team determined there is no procedure to delineate the use of green tags for control of QA Class 1, Commercial Grade, and EQ material, parts, and components.
Lack of a procedure and present practice may allow non-qualified parts or components tc, be used in safety / quality related equipment thereby jeopardizing the functionality of the components or equipment.
This situation may compromise Rancho Seco's commitment to 10 CFR 50, Appendix B.
Vendor manuals are not programmatically controlled.
Numerous examples of vendor manual inaccuracies and inadequate use were identified.
Nuclear Engineering is not reviewing vendor manual correspondence for applicability prior to inclusion into vendor manuals.
Additional concern is that vendor correspondence (letters, drawings, charts, etc.) is sent to individuals unaware of their responsibilities with respect to revising and maintaining vendor manuals.
4.3 The Emergency Operating Procedures (EOP) requires Pressurizer Auxiliary Spray line flow that is significantly higher than flow that can be physically achieved (calculated) and at stated in Design Bases Document (DBD).
The DBD fcr SIM requires a flow to Auxiliary Pressurizer Spray of 40 gpm, thru an orifice to ensure approximate 40 gpm flow, with a pressure drop of 2490 psig.
The DBD also requires cooldown of the pressurizer with use of the Auxiliary Spray from 325 psig to atmospheric pressure and decrease temperature from 428'F to 212*F in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, to be accomplished with 40 gpm flow rate.
The EOPs, however, require flow rate in the Auxiliary Spray line of 105 gpm.
Pressurizer cooldown would be much faster than required in the DBD.
Also, calculations indicate that 105 gpm flowrate through an orifice sized for 40 gpm could not be achieved.
[RI-135]
5.0 SPECIFIC CONCERNS A list of the specific concerns the Expanded ASRTP team believes are i
new concerns not previously identified for resolution follows:
5.1 Acknowledged (Valid) Concerns j
i 5.1.1 A review of data from the four identical High Pressure i
Injection (HPI) M0 valve operators found discrepancies in settings and setpoints.
[RI-119]
5.1.2 Inadequacies with MO valve procedure EH.117A include:
Instructions for torque and limit settings Testir, of component functions Setnng and testing setpoints Valve operator interchange Engineering involvement
[RI-121]
5.1.3 Commercial grade items bought to date and used in l
l safety-related equipment may compromise the EQ life of the equipment because a formal Commercial Dedication Program is not in place.
[RI-124]
5.1.4 E0P requirement for Pressurizer Auxiliary Spray rate may cause pressurizer cooldown significantly different from DBD requirement.
[RI-135]
5.1.5 A program is not implemented that provides adequate control of Vendor Technical Manuals.
ERI-151]
5.1.6 Adequate tests or calculations are not in place to ensure that Makeup and HPI pumps have adequate NPSH when taking i
suction from DHR System (PIGGY BACK MODE).
[RI-153) i 5.2 Open (Potential) Concerns l
5.2.1 None I
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6.0 ATTACHMENTS 6.1 List of Documents Reviewed
- 6. 2 Status of RIs 6.3 Detailed Observations - Requests for Information i
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LIST OF DOCUMENTS REVIEWED 1)
Updated Safety Analysis Report, Sections: 5.2, 9.2, 12.2.11, 14.1.3, 14.2.2.5.
2)
Design Bases Document, NEAP 5471, Rev. 1 SIM 3)
System Status Report SIM Rev. 1 l
4)
Code of Federal Regulations,10CFR21,10CFR50 5)
Systems Training Manuals:
Chapter :
5 Hake-up and Purification j
27 Emergency Core Cooling 35 Safety Features Actuation System 6)
Training Course Outline:
00-24I0300 00-21I2500 7)
Technical Specifications, 3.12, 3.2.2.1, 3.3.1.2, 3.5. 1.1, 3.6.1, 4.1.1, 4.5.1.1, 4.2, 4.4, 4.14 8)
Emergency Operating Procedures:
E.01 Rev. 2 E.02 Rev. 5 E.03 Rev. 3 E.04 Rev. 7 E.05 Rev. 9 E.06 Rev. 7 E.07 Rev. 5 CP 101 Rev. 4 i
CP 102 Rev. 2 CP 103 Rev. 3 CP 104 Rev. 4 CP 105 Rev. 3 i
9)
Plant Operating Procedures:
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A.1 Rev. 21 Reactor Coolant System j
A.2 Rev. 16 Reactor Coolant Pump System j
A.3 Rev. 21 Pressurizer and PRT System i
A.4 Rev. 15 Core Flood System A.15 Rev. 31 Make Up, Purification, and Lvedown System 8.2 Plant Heat-up and Start-up 8.4 Plant Shutdown and Cooldown l
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I ATTACHMENT 6.1 i
LIST OF DOCUMENTS REVIEWED (Continued) 10)
Casualty Procedures:
C.9 Rev. 3 Loss of Reactor Coolant Make-up/ Letdown C.11 Rev. 2 Pressurizer System Failures C.15 Rev. 10 Loss of NNI 'x' Power C.20 Rev. 9 Loss of CCW C.41 Rev. 1 Recovery from SFAS Actuation C.95 Rev. O Loss of Offsite Power 11)
Administrative Procedures:
AP 23 Conduct of the Operations Division AP 46 Rev. 4 AP 46 Rev. 5 AP 102 Rev. 9 Process Standards
- 12) Maintenance Procedures:
LDAP 003 Rev. 0 USAR Revision Control 14)
Nuclear Engineering Procedures:
NEP4109 NEP4202 NEP4203 15)
In Service Testing Program Plan 16)
Generic Letter 83-28 " Vendor Interface Program" 17)
QA Audit Report 0-734 " Technical Specification ECCS" 9-5-85
- 18) QA Audit Report 87-038 " Technical Specification Current Outage" 6-19-87 ATTACHMENT 6.1.
LIST OF DOCUMENTS REVIEHED (Continued) 1 19)
NRC IE Bulletins and Notices:
j IEN 82-35 Memo EQC 83-124 Response to IEN 82-35 I
IEN 83-72 Environmental Qualification Testing Experience IEN 82-52 Equipment Environmental Qualification Testing l
Experience j
i IEN 82-17 Over pressurization of Reactor Coolant System IEN 81-15 Degradation of Automatic ECCS Actuation capability by Isolation of Instrument lines IEN 80-44 Actuation of ECCS in the Recirculation mode while
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in Hot Shutdown i
IEN 82-45 PWR Low Temperature Overpressure Protection l
IEN 87-01 RHR Valve Misalignment Causes Degradation of ECCS in PHRs I
IEN 82-19 Loss of High Head Safety Injection Emergency j
Boration and RC Hake-up Capability 1
IEN 81-31 Failure of Safety Injection Valves to Operate Against Differential Pressure IEN 82-20 Check Valve Problems IEN 82-35 Failure of Three Check Valves on High Injection i
Lines to Pass Flow l
IEN 81-35 Check Valve Failures j
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- 20) Calculation Z-SIM-M0321 1
- 21) Vendor Technical Manual N7.02-22 Bryon Jackson Pump Division, Instruction Manual for P236, P238A, P2388 j
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- 22) ASME Code NC 3133
- 23) Surveillance Procedures:
SP 214.02 j
SP 214.03 SP 203.01 A,B SP 200.09 l
SP 203.02A I
SP 203.028 l
SP 203.02C I
- 23) Special Test Procedures:
STP-1054 SIM MOVATS Test l
- 25) Drawings P& ids:
520 Sheet 1, 2 M521 Sheet 1,, 2 and 3 i
M522 M526 M525 M570 M544 M560 Sheet 1 l
ATTACHMENT 6.1
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l LIST OF DOCUMENTS REVIEWED (Continued) j 26)
Elementary Diagrams:
E 1012 E 203 sh 50E, F, G, H D 203 sh 54
- 27) Non Conformance Reports (NCR):
S 00621 S 03547 l
S 03827 S 03867 l
S 04152 S 04772 S 05241 S 05684 28)
ECN R-0089 ECN A-1119 ECN A-1240 ECN A-1770 ECN A-5260 1
- 29) Work Request #:
14835 45815 43202 63919 100610 111168 110739 100622 55018 63911 17002 104979 109274 69184 84867 14100 15909 l
76323 I
71209 17039 i
ATTACHMENT 6.1 j
LIST OF DOCUMENTS REVIEWED (Continued) 30)
Calculations:
Z-EDS-E0005, Rev. 0 - Motor Starting Under Normal Cold Start and Hot Restart System Operation Z-EDS-E0006, Rev. 1 - Voltage Regulation Studies - Normal Steady State Operation Z-EDS-E0007, Rev. 0 - Emergency System Operation E-0711, Rev. 0
- Short Circuit Study of Auxiliary Power Distribution System (dated 07-21-87, under review by SMUD) l Z-FWS.E0178
31)
NEP 5104.1 - Electrical System Design Parameters NEP 5104.2 - Selection and Sizing of Power and Control Cables NEP 5204.38 - Design of DC Power Circuits NEP 5204.39 - Calculation for AC and DC Control Circuits 32)
Procedure:
EM.117A - Testing of Limitorque Motor Operated Valves Using H0 VATS I
- 33) Drawings:
E-1012, Sh.08 - Motor Operated Valves Data E-1012, Sh 109 -
E-1012, Sh 110 -
E-1012, Sh 111 -
E-10ll, Sh 2, Rev. 10 - Load Design Data Sheet l
E-1011, Sh 8, Rev. 9 E-10ll, Sh 72, Rev. 7 -
E-1011, Sh 74, Rev. 10 -
E-1011, Sh 75, Rev. 10 -
E-305, Sh 3 - Hiring Diagram 125V DC System E-203, Sh 50E - Elementary Diagram E-203, Sh 50F -
E-203, Sh 50G -
l E-203, Sh 50H -
l E-203, Sh 37A -
E-203, Sh 37 E-203, Sh 4A -
E-203, Sh 4 E-203, Sh 2A E-203, Sh 2 E-203, Sh 22 -
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E-203, Sh 74 ATTACHMENT 6.1 l
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STATUS OF RIs.2 provides RI status as of this report date.
An RI is considered closed if the Team Leader was convinced a potential concern was not valid or not significant enough to be an RI.
An RI would also be
)
closed if requested information was provided.
All other RIs are open.
Acknowledged RIs are open RIs that have been accepted as valid by the responsible organization and have been stated as concerns in Section 5.0.
RI NUMBER STATUS 106 CLOSED 111 CLOSED 118 CLOSED 119 ACKNOWLEDGED 120 CLOSED 121 ACKNOWLEDGED 124 ACKNOWLEDGED 135 ACKNOWLEDGED 136 CLOSED 137 CLOSED 151 ACKNOWLEDGED 153 ACKNOWLEDGED l
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1 ATTACHMENT 6.2 ;
1
DETAILED OBSERVATIONS - RE0 VEIT FOR INFORMATION During an evaluation, all potential concerns are documented on Request for Information sheets.(RIs) that are sent to the responsible organization to receive their input concerning the potential concern, RIs are also used to request information that the EASRTP team is having difficulty obtaining.
These RIs are considered draf ts throughout the entire evaluation until they become part of the report.
Responsible organizations can accept the potential concern as valid or they may disagree with the potential concern.
If they disagree, they can submit information that convinces the EASRTP team members that the potential concern is not valid, or they may redirect the EASRTP members to better focus the concern.
RIs developed during the system evaluation comprise this section of the i
. report.
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I ATTACHMENT 6.3 l A
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REQUEST FOR INFORMATION (RI)
RI NO:
106 SYSTEM CODE:
SIM ISSUE DATE:
08-05-87
SUBJECT:
SMALL BREAK LOCA HPI OPERATION DEPARTMENT:
NUCLEAR LICENSING COORDINATOR:
JERRY DELEZINSKI/3909 i
TEAM LEADER:
KEITH PRINCE /3851 POTENTIAL CONCERN /0UESTION:
Failure of D batteries in conjunction with a break at the HPI nozzles will require operator action to fully open the HPI valves to provide sufficient flow for this accident condition.
The Design Bases Document (DBD) for SIM System provides minimum HPI flow for small break LOCA (page 27 of 69) which can be achieved by 2 HPI Pumps following through 2 HPI lines and by opening 2 HPI Valves manually beyond l
their pre-throttled SFAS open position.
Also, Emergency Operating Procedure E02 states, "If pressurizer level is still decreasing, open additional HPI Injection valves (Rule 2)."
Letters from SMUD to the USNRC dated July 20, November 22, and December 4, all 1978, addressed actions and modifications incorporated so as not to require operator action in the event of a small break LOCA.
The requi. red flow can be achieved without operator action only up to 10 l
minutes after initiation after which time operator action is required.
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o This RI is closed.
Concer satisfied by Licensing, l
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REQUEST FOR INFORMATION (RT)
RI NO:
111 SYSTEM CODE:
GENERIC ISSUE DATE:
08-06-87
SUBJECT:
AUDITS DEPARTMENT:
OUALITY ASSURANCE COORDINATOR:
DAVE HALONE/3851 TEAM LEADER:
KEITH PRINCE /3851 POTENTIAL CONCERN /0UESTION:
Paragraph 6.5.2.8 of the Rancho Seco Technical Specifications requires that audits shall encompass:
"The conformance of facility operations to all provisions contained within the Technical Specifications and applicable license conditions at least once a year..."
Discussions with QA Internal Audit Personnel revealed that this requirement could not be met because of manpower constraints and has been previously identified in Internal Audit Report No.56-002.
- However, resolution to this deviation has not been provided by QA Management according to the Internal Audit Group.
4 This condition was noted when determining if the QA Audit Group has j
performed as audit to ascertain compliance to Technical Specification I
requirement 4.5.3, certinent to the decay heat removal system.
The last previous audit to a Jress this was performed December 2 thru 13, 1985 (Audit 0-770).
o This RI is closed.
This item was identified previous to EASRTP.
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REQUEST FOR INFORMATION (RI)
REVISION 1 RI NO:
118 SYSTEM CODE:
SIM ISSUE DATE:
08-17-87
SUBJECT:
USAR DISCREPANCY DEPARTMENT:
LICENSING /NE COORDINATOR:
JERRY DELEZINSKI/3909 TEAM LEADER:
KEITH PRINCE /3851 POTENTIAL CONCERN /0UESTION:
The operation of a SIM Safety Features valve is not consistent with valve description in the USAR.
USAR Chapter 5, Section 5.2 and Table 5.2-2 states that the Reactor Coolant Pump seal water supply Motor Operated Safety Features Valve (SFV-23616) has a post accident position of " closed." However, Engineering Change Notice (ECN) A-1240 removed the safety features signal from valve SFV-23616.
A new interlock was provided to close SFV-23616 only when all four Reactor Coolant Pumps are not running and a low seal injection flow condition exists.
This change was completed and signed off on 08-16-76.
Elementary diagram E-203, Sh. 54, Rev. 4, reflects the removal of the safety features signal.
licensing Department Administrative Procedure LAP-0003 requires the open and closed ECNs to be reviewed from a list available from Site Documer.t Control, and for changes to the USAR to be done annually.
The discrepancy between the description of SFV-23616 in the USAR and the actual interlock function should have been corrected at a reasonable time after 08-16-76 when the ECN was closed.
o This RI is closed.
Licensing requested USAR to be updated.
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a
REQUEST FOR INFORMATION (RI)
REVISION 1 RI NO:
119 SYSTEM CODE:
SIM ISSUE DATE:
08-12-87
SUBJECT:
MOTOR OPERATED VALVE DATA SHEETS E-1012 DEPARTMENT:
NUCLEAR ENGINEERING COORDINATOR:
R. LAWRENCE TEAM LEADER:
KEITH PRINCE /3851 j
POTENTIAL CONCERN /0UESTION:
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Potential discrepancies in documentation of MOV Operator settings and setpoints.
There are 4 HP Injection Valves.
These are:
Valve #
Paramettr_.Sfj;coint Ehjell SFV-23809 E-1012 Sheet 108 SFV-23810 E-1012 Sheet 109 SFV-23811 E-1012 Sheet 110 SFV-23812 E-1012 Sheet 111 These 4 valve operators are identical with regard to make, model, function environment, etc.
a)
E1012 sheet 111 indicates 125 VAC.
The remaining 3 sheets indicate 125 VQC.
b)
Note 6(a) is included in sheets 109 and 111 only, where torque switch cut off coast down thrust is limited to 15,400 lbs (110%
of operator capacity).
This limitation is not provided on the other valves.
c)
Sheet 110 has considerable difference '.n the limit switch setting (% of disc movement) Actual from the rest of 3 valves.
d)
Strcke lengths are different for all the valves.
e)
Maximum hand wheel torque design is 60 ft lbs in sheet 109 and 44 in the other 3 sheets.
f)
The design thrust AT delta P is 11,729 in sheet 110 and 11,734 in the other 3 sheets.
g)
Maximum stalled thrust is 65,426 lbs in sheet 110 where as 65,708 on the other 3 sheets.
h)
SFV-23811 stroke time is recorded as 12 seconds.
3/fnaining valves required 14 seconds.
Also, discrepancies were found between E-1012 (Sheet 110) and EH.ll7A-18 for Valve SFV-23811.
E-1012 EH.ll7A-18 Max Design Running Current 9.4 amps 9.1 amps Horsepower 1.083 1.033 Open Torque Switch Setting 1% - 2%
Blank Close Torque Switch Setting 1% - 2%
Blank \\.
REQUE3T FOR INFORMATION (RI)'
REVISION 1 RI NO:
120 SYSTEM CODE:
SIM ISSUE DATE:
08-14-87
SUBJECT:
MOV TESTING PROGRAM DEPARTMENT:
NUCLEAR ENGINEERING COORDINATOR:
R. LAWRENCE /4365 TEAM LEADER:
KEITH PRINCE /3851 POTENTIAL CONCERN /0UESTION:
During review of EM-ll7A data sheets for valve SFV-23811 and the respective E1012 sheet 110, discrepancies were noted.
Examples are listed below:
E-lD12 EM117A-18 a)
Max Design Running Current 9.4 AMPS 9.1 AMPS b)
Horsepo*wer 1.083 1.033 c)
Open Torque Switch Setting 1 3/4-2 3/4 blank d)
Close Torque Switch Setting 1 3/4-2 3/4 blank The valve SFV-23811 was set and adjusted to the Work Request #121711 Rev.
1.
Please identify which document was used in the field?
EM.117A was signed off by the MOVAT Testing Group and E1012 was signed off by MOVAT Design Engineer.
Please advise who reviews and approves the DCNs for MOVATS and which group verifies that all information contained in the data sheets are correct?
o This RI is closed.
The contents were combined with RI 119.
REQUEST FOR INFORMATION (RI)
RI NO:
121 SYSTEM CODE:
SIM ISSUE DATE:
08-12-87 l
SUBJECT:
MOTOR OPERATED VALVE PROCEDURES l
DEPARTMENT:
NUCLEAR ENGINEERING COORDINATOR:
R. LAWRENGZ3849 TEAM LEADER:
KEITH PRINCE /3851 POTENTIAL CONCERN /0UESTION:
Listed below are apparent inadequacies in motor operated valve procedure EM-117A.
Discussions with MOVATS group, Nuclear Operations and maintenance people plus additional review of EM-117A did not clear the following concerns:
1)
There appears to be a lack of proper instructions in procedure EM-ll7A, Revision 3, with regard to setting of torque and limit switches.
All functions of limit switch settings are not addressed in the procedure, nor are all functions being tested. (e.g., SFAS initiation pre-throttle positions) 2)
The procedure does not address settings or setpoints and testing of external limit switches.
3)
The procedure does not cover the adjustment or verification of valve local position indications.
4)
There is no Nuclear Engineering (NE) Procedure covering the MOVAT activities, including instructions to complete the E-1012 series for motor operated valves.
Si EM-117A does not address cannibalizing parts and swapping operators for use on different MOVs.
Nuclear Engineering has no procedure to qualify review the EQ, SEQ and other Technical Parameters of the operators for their intended use.
6)
It appears that MOVAT qualified Maintenance Engineering Staff accepts the MOVAT test data sheets.
There is no independent review by Systems / Design Engineering group over and above a second level reviewer from the MOVAT group.
RI M1 (Continued) 7)
Procedure EM-Il7, paragraph 3.1.7 states that torque switch setting will be per E-1012 or at direction of Nuclear Engineering.
- However, paragraph 7.7 Item i states setting as determined by Electr': cal Maintenance Engineering.
These statements appear to contradict.
8)
Procedure EM-117A, paragraph 6.1.5.6 states that open torque bypass switch to open at the percentage specified on the associated E-1012 drawings.
However, E-1012 does not have a line that indicates the percentage.
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REQUEST FOR INFORMATION (RI)
REVISION 1 RI N0:
124 SYSTEM CODE:
GENERIC ISSUE DATE:
08-19-87
SUBJECT:
PROCUREMENT AND DEDICATION OF COMMERCIAL GRADE ITEMS DEPARTMENT:
ENGINEERING C0ORDINATOR: DAVE MALONE/ JIM DARKE TEAM LEADER:
KEITH PRINCE /385_1 POTENTIAL CONCERN /0UESTION:
There is no program in place for dedication of commercial grade items.
The team is concerned that procurement and installation of commercial grade parts for safety-related and EQ related equipment and components without a formal program may invalidate the quality life of some components and equipment.
The team is aware that a procedure (RSAP-0706, Rev. 0) to delineate requirements for " Dedication of Commercial Grade Items" is in the approval process.
Commercial grade items installed to date, however, cause concern because the team cannot determine that necessary actions were taken or documented to validate commercial grade items. These include:
]
Nuclear Engineering has not specified receipt sample plan, inspection criteria, nor acceptance criteria for receiving commercial parts.
A Commercial Vendor List could not be found.
1 Engineering /QA source inspections of Commercial Vendors were not performed.
No training of appropriate personnel on Commercial Dedication has been done.
Requirements for mild environment EQ replacement parts are not specified.
There is no procedural control of green tags used for control of commercial grade, QA Class 1, and EQ qualified materials, parts and components.
Rancho Seco's commitment 10 CFR 50, Appendix B (Identification and Control of Materials, Parts, and Components) may be compromised.
Without such control, it is possible for Non-Class 1 or non-EQ components to be placed into a safety-related system and jeopardize the ability of the system to perform its intended function.
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REQUEST FOR INFORMATION (RI)
RI NO:
135 SYSTEM CODE:
SIM ISSUE DATE:
08-18-87
SUBJECT:
PRESSURIZER AUXILIARY SPRAY DOCUMENT DISCREPANCY DEPARTMENT:
NUCLEAR ENGINEERING COORDINATOR:
R. LAWRENCE /4365 NUCLEAR OPERATIONS R. MACIAS/4589 T i a'. LEADER:
KEITH PRINCE /3851 POTENTIAL CONCERN /0VESTION:
There is discrepancy between Design Bases Document (DBD) and Emergency Operating Procedures that could affect Plant cooldown rate.
Emergency operations procedure E0P CP101 (Section 13.3) states minimum requirement of flow of 105 gpm to be established by throttling SFV-23810 when using pressurizer spray and SFV-23801 closed.
The Design Bases document for SIM system, however, states (Section 6.2.1) the design bases for flow orifice FP26045 as 40 gpm with a pressure drop of 2490 psig.
This design is based on SMUD Calculation #Z-SIM-M0188.
Also, the Design Bases document Section 6.1 (24) has the requirement for use of pressurizer auxiliary spray to cooldown the pressurizer from 325 psig to 0 psig and decrease the temperature from 428 degrees F to 212 degrees F in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> which is accomplished by 4_Q gpm through the auxiliary spray, not 105 gpm.
A flow rate of 105 gpm as opposed to design bases 40 gpm would greatly affect pressurizer cooldown rate.
Calculations indicate that it would not be possible to pass 105 gpm through an orifice designed for 40 gpm. L
l REQUEST FOR INFORMATION (RI) l l
i i
RI NO:
136 SYSTEM CODE:
SIM ISSUE DATE:
08-18-87 i
SUBJECT:
PRESSURIZER AUXILIARY SPRAY ORIFICE SIZING DEPARTMENT: _ SYSTEMS ENGINEERING COORDINATOR:
J. ITTNER/4153 TEAM LEADER:
KEITH PRINCE /3851 POTENTIAL CONCERN /0UESIIQH:
Inability to determine accurate flow rate through Pressurizer Auxiliary l
1 Spray could affect Pressurizer cooldown rate.
The auxiliary spray is designed to provide cooling water to the Pressurizer spray when the RCPs are not in operation.
Flow to the spray is monitored by flow transmitter FT-23806.
SIM SSR (Problem 27) states that FT-23806 is not reliable at low flow rates.
Flow orifice FP-26045 should be 40 gpm by Design Bases.
Without l
a reliable flow transmitter, actual flow or verification of orifice condition cannot be determined.
o This RI is closed.
This problem was identified previous to EASRTP.
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REQUEST FOR INFORMATION (RI)
RI NO:
137 SYSTEM CODE:
SIM ISSUE DATE:
08-17-87
SUBJECT:
TESTING OF SEAL INJECTION FLOW CAPABILITIES DEPARTMENT:
SYSTEM _1 ENGINEERING COORDINATOR:
J. ITTNER/4153 TEAM LEADER:
KEITH PRINCE /3851 POTENTIAL CONCERN /0UESTIQfi:
There is a valve function listed in the Design Bases Document (DBD) that is not verified to be capable of tne described function.
The DBD section 6.1 -- item 25 describes the system function and reason as follows:
Design seal injection flow rate to each RC pump was 8 gpm with capability of supplying 16 gpm without adjustment of the manual needle valve which is pre-throttled.
RC pumps B, C, and D have a newer seal design which uses 9.5 gpm seal injection.
Design flow rate selected when RC pump design was changed to delete throttle bushing and use two mechanical seals with a controlled bleedoff.
For a loss of component cooling water, 16 gpm was believed to be potentially necessary to maintain acceptable oearing temperature.
o There appears to be no test to verify this function.
Plant Operating Procedure A.15 section 4.2.21
.28 provides the steps necessary to establish balanced seal injection flow each time the SIM system is started up.
This may prevent seal injection from being increased as stated above, since varying condition can exist when setting up the flow rates through the needle valves.
o Casualty Procedure C.20 section 2.4, " Complete loss of CCH flow",
has the operator reduce seal injection flow to 12 gpm (3 gpm/ pump) using PV-23606, RC Pump Seal Injection Flow Controller.
The reason for reducing seal inject flow to a minimum is to prevent overfilling the pressurizer since there is no letdown capability with loss of CCH.
This is in contradiction to the reason given for increasing seal injection flow in the DBD.
o Is the capability of increasing seal injection as described in the DBD an actual requirement for equipment protection?
If so, how te it verified to have this capability? u_ _---
o Why do the Plant Operating Procedures instruct the operator to reduce seal injection flow, when the DBD states the need for increasing seal injection flow?
o What analysis is done for potential damage to the reactor coolant pump bearing due to increasing temperature from reduced seal injection flow with a loss of CCW?
I o
This RI is closed.
No real problem found upon further as>essment.
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l REQUEST FOR INFORMATION (RI)
RI NO:
151 SYSTEM CODE:
GENERIC ISSUE DATE:
_Q3-18-87
SUBJECT:
CONTROL OF VENDOR TECHNICAL MANUAL DEPARTMENT:
SITE DQCUMENT CONTR_0L CONFIGURATION CONTROL COORDINATOR:
SKIP WOOD /4728 RON LAWRENCE (DICK HARLOW)/4365 TEAM LEADER:
KEITH PRINCE /3851 POTENTIAL CONCERN /0VESTION:
Vendor Technical Manuals are not being controlled in accordance with AP.46; training of personnel in vendor manual control has not been implemented; and clear cut responsibility for control of vendor manuals has not been accepted.
The concerns include:
The responsibilities imposed upon Nuclear Engineering Document Control (NEDC) as defined in Section 6.0 of AP.46 are being implemented by personnel in the Technical Library..1 of AP.46, " Technical Manual Review Sheet," is not being utilized as required by Section 6.2.
It should be noted, however, that the review sheet currently used requires essentially the same information as Enclosure 8.1.
Material (letters, drawings, etc.) from vendors that should be included in the manuals is being received by persons not aware of this and the material is being stored in files, or perhaps discarded.
AP.46 does not reference AP.42, " Maintenance Information Management e
System" (MIMS).
The MIMS Program contains the Master Equipment List (MEL), which in part provides reference to Manufacturer's Instruction Book numbers.
Distribution of technical manuals for review and approval is done by the Technical Library, not NEDC.
A training program for those persons that use vendor manuals has not been implemented. l
REQUEST FOR INFORMATION (RI)
RI NO:
153 SYSTEM CODE:
SIM IS$l'E DATE:
08-18-87
SUBJECT:
MAKE-UP/HPI PUMP NPSH DEPARTMENT:
NUCLEAR ENGINEERING COORDINATOR:
R. LAWRENCE /4365 TEAM LEADER:
KEITH PRINCE /3851 PQTENTIAL CONCERN /0UESTION:
Assurance is not provided that the HPI pumps will have adequate NPSH for any conditions that may be encountered under Emergency Operating Procedures.
There are no calculations to document that the Make-up and HPI pumps will have adequate NPSH when taking suction from the Decay Heat Removal System
(" piggyback mode").
This RI supercedes RI-71.
j The original test, TP203-4, demonstrated that plenty of NPSH margin is available when delivering 528 gpm to one pump with the decay heat system at near-shutoff head.
This is a nonconservative case.
By following E0P guidelines (Rule 2 and CP.101), it is possible to get nearly twice as much flow through the line with the Makeup Pump and HPI Pump both operating off the common suction header.
In the long term cooling mode (CP.101 Steps 12, 13) an HPI Pump will i
be operated in the " piggyback" mode with the RCS at atmospheric pressure.
Since the " piggyback" line branches off downstream of the i
throttle valve (SFV-26039/SFV-26040), there is minimal pressure available for HPI pump NPSH.
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