ML20239A084
| ML20239A084 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 08/27/1987 |
| From: | Akins M, Croley B, Humenansky D SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20238F564 | List: |
| References | |
| NUDOCS 8709170069 | |
| Download: ML20239A084 (26) | |
Text
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EXPANDED AUGMENTED SYSTEM REVIEW AND TEST PROGRAM (EXPANDED ASRTP)
EVALUATION-0F THE NUCLEAR SERVICE COOLING HATER SYSTEM SUBMITTED BY:.
DATE: d d 7
M. J. AKINS U
TEAM LEADER CONCURRENCE:
Mi
%W DATE:~
" lL hl EXPANDED ASRTP PROGRAM M(NAGER
' DAVID HUMElfANSKY d
/.
/27/
CONCURRENCE:
DATE:
DIRECTOR, NUCLEAR [TECHNICALSERVICES-BOB CROLEY I
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TABLE OF CONTENTS Pace Number
1.0 INTRODUCTION
3 2.0 PURPOSE 4
3.0 SCOPE 5
4.0 OVERALL RESULTS AND CONCLUSIONS 6
5.0 SPECIFTC CONCERNS 8
l 5.1 Acknowledged (Valid) Concerns 8
5.2 Open (Potential) Concerns 8
6.0 ATTACHMENTS 9
6.1 List of Documents Reviewed 10 6.2 Status of RIs 13 6.3 Detailed Observation - Requests for Information 14
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1 _ ______________ ___________ _ ______
EXPANDED AUGMENTED SYSTEM REVIEW AND TEST PROGRAM EVALUATION OF THE NUCLEAR SERVICE COOLING HATER SYSTEM i.0 INTRODUCTION 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 of ongoing 9rograms 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 implemented on 33 selected systems and an ia-depth review of the engineering, modification, maintenance, operations, surveillance, inservice testing, and quality programs.
It also conducts a review, 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 with appropriate backgrounds to evaluate the operations, maintenance, ' engineering, and design functional areas.
Independence, perspective, and industry 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 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 order to enh'ance the evaluation.
Each team prepares a report for i
l each completed selected system evaluated.
This report is for the Nuclear Service Cooling Hater (NSCH) System.
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l 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 ensurin]
safety during plant operation after restart.
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3.0 SCOPE To' accomplish the first objective, the Auxiliary Systems team evaluated the NSCH system to determine whether:
1.
The system was capable of performing the safety functions required by its design bases.
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 adequate 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 2.
Maintenance 3.
Operations and Training 4.
Surveillance and Inservice Testing
- 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 NRH evaluation report and the NSCH 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 NSCH system included a review of pertinent portions of support systems that must be functional in order for the NSCH system to meet its design objectives.
I 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 brackets after each individual summary or concern refer to the corresponding RIs in Attachment 6.3.
The conclusion reached from the review of the Nuclear Service Cooling Water System is, that there are enough inconsistencies in the design specifications, calculations and surveillance procedures to raise a concern about the ability of'the system to meet its intended functional requirements.
This inability is closely coupled with the inconsistencies found in the NRH system.
- However, correcting the NRW system will not resolve the NSCH problems.
4.1.
Nuclear Service Cooling Hater System Heat Removal The primary purpose of the NSCH system is to remove heat from the reactor coolant system during cooldown via the Decay Heat Removal (DHR) Coolers, and from Containment during post-LOCA via the DHR Coolers and the Reactor Building Emergency Air Coolers.
Based upon discussions with B&W, and a review of the USAR, and various manufacturer
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drawings, the DHR coolers were specifically designed to l
remove 125 x E6 BTU /hr for post-LOCA conditions and 33.75 x E6 BTU /hr for shutdown conditions with a NSCH inlet temperature of 95'F.
l A review of the system calculations as well as the Nuclear Service Raw Water system capabilities indicate that the NSCH system cannot provide water at or near the 95'F design requirement.
This was clearly identified in both the 1969 Calculation Z-NSH-H0251, and the NSCH Heat Exchanger data sheet.
Based on these findings, the DHR Coolers are not able to provide their required cooling at the design basis accident and normal shutdown conditions.
[RI 128] [RI 107]
This problem is complicated by the fact that the flow rate on the NSCH side of the DHR coolers is to be maintained l
below 3750 gpm (as a result of B&W letter to Bechtel, dated 7 Mar 73) to prevent tube damage.
Based upon recent surveillance (08-20-87) data, this flow rate limit-through j
the Coolers may have been exceeded.
[RI 104]
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i OVERALL RESULTS AND CONCLUSIONS (Continued) l 4.2 Potential Loss of System Capability Based upon the commitments in the USAR (section 9.4.2.1),
the radiation monitoring in the NSCW system is to be continuous.
Also, the non-safety, non-seismic class piping l
connected to the safety class system is to be isolated via a i
i closed safety-class valve.
The P&ID (dwg M-544) and several isometric drawings show that the piping to and from the radiation monitors are non-seismic.
The P&ID and the operational valve lineup show that the isolation valves are normally open.
Therefore, the system as designed cannot be isolated per the USAR commitment for continuous monitoring.
l As presently configured, a seismic event could cause both loop A & B monitor piping to fail resulting in a system draindown that trips both pumps on low-low water level in the surge tanks.
This is complicated by the fact that neither loop has an independent seismic class makeup capability.
This would tend to increase the duration during which the system would be inoperable.
[RI 105]
Refurbishment of NSCW pump B has been completed.
A review l
of work requests has indicated that the casing bolts were l
torqued to a value three times the manufacturer's i
requirements.
There was no apparent review or discussion I
with the manufacturer about the reasons for exceeding the limit.
The vendor / representative indicated the cast iron casing threads could be stripped by the excessive torque l
that could lead to a complete failure of the casing.
The pump casing gasket was replaced with a gasket which was twice as thick as specified by the pump manufacturer.
The vendor representative stated that the stationary wear rings will probably wear at an accelerated rate and cause l
premature degradation of the pump.
[RI 132]
The work request used to refurbish Pump B required the QC representative to initial a " Hold" point signifying his review of all installed materials.
The documents lacked the i
required initialing by QC on several items including the
" Hold" point.
[RI 141]
4.3 Testing j
Based upon a review of available testing documents and the USAR, no evidence could be found in documents that the NSCW j
system was pressure tested in accordance with the j
i requirements specified in ASME Section XI, paragraph IHD j
i 5000.
[RI 109]
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l-5.0 SPECIFIC CONCERNS A list of the specific concerns the Expanded ASRTP team believes are concerns not previously identified for resolution follows:
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5.1 Acknowledged (Valid) Concerns 5.1.1 A failure of the non-safety class piping to the NSCH radiatic monitors could render both NSCH loops inoperable.
[RI 105]
l 5.1.2 Calculations did not balance parallel flow rates simultaneous; through NSCH system components.
This same problem was founc the calculations for the Nuclear Service Raw Hater System.
[RI 107]
5.1.3 The Nuclear Service Cooling Hater System is not capable of supplying cooling water at a temperature necessary to meet normal shutdown and safety-related design requirements.
[R:
5.1.4 The NSCH system has not been pressure tested in accordance l
the Inservice Inspection requirements specified in ASME Secr ~
I XI.
[RI 109]
5.1.5 The flow rate through the NSCH (shell) side of the Decay Hea:
I Removal heat exchanger may have exceeded the maximum recomme-to prevent tube damage.
[RI 104]
f 5.2 Open (Potential) Concerns 5.2.1 The integrity of both NSCH pumps may have been compromised du to the work that was recently performed in accordance with W Requests #0135203-0/1 and 0133421-0.
[RI 132]
5.2.2 It is not possible to perform a verification of the limit anc.
torque switch settings for MOVs derived from Nuclear Engineer' j
calculation Z-ZZZ-M2262.
[RI 140]
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6.0 ATTACHMENTS 6.1 List of Documents Reviewed 6.2 Status of RIs 6.3 Detailed Observations - Request for Information 1
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1 LIST OF_QQCUMENTS REVIEWED ERi S-3709, S-3178, A-4905, C-6499 S-5257, S-4565, 4913 S-5523, S-4726, S-4969 S-5709, S-4727, S-5023 S-4552, 5242, S-5425 6805, Rev 0, 1, 2, S-5464 6816, S-5499, S-6103 6817, S-5584, S-5417 6818, S-5645, S-6908 1
Work Reauesti 076131, 097504, 097997, 099554, 099555, 098588, 098589, 097593, 1006; 105788, 105886, 105887, 101797, 101027, 107028, 107031, 107032, 1070.7 107906, 108533, 110863, 112709, 111359. 113983, 115340, 115720, 1158' 116006, 116406, 116539, 116540, 116796, 117254, 119468, 098730, 0779' 092880, 099956, 111840, 098768, 105120, 114587, 114611, 114900, 1146' 115333, 115719, 115760, 115761, 115809, 115864, 116232, 116525, 1170; 117255, 117256, 117257, 117258, 121770, 66550, 60027, 57222, 75196, 45781, 46814, 0135203, 105888, 133421, 116535, 117689, 115720, 10984:
107026, 107029, 107030, 107034, 132070 Drawinas E-203, Sh 43, Rev 12, E-105 Sh 9A, Rev 9, 10, 11 E-208, Sh 20A, Rev 4 E-105 Sh 13, Rev 23 l
E-342, Sh 66 Rev 2, 4, 6, 10 E 1012, Sh 171 Isometric Drawinas 48200-1"-DB3 Rev 2 48201-1"-D83 Rev 2 48210-1"-0B3 Rev 0 48211-1"-0B3 Rev 0 48220-1"-DB3 Rev 3 48221-1"-083 Rev 3 48600-1"-HE 48602-1"-HE 48620-1"-HE 48621-1"-HE l
48690-1"-HE 48692-1"-HE 48225-1"-HE 48226-1"-HE 48470-1"-DB sheet 1, lA, 2, 3, 4, 5 50050-10"-HE sheet 1, lA, 2 50054-10"-HE ATTACHMENT 6.1 L___________-____________________
LIST OF DOCUMENTS REVIEHED System Trainina Manual Chapter #37, #30 STP 10848, Rev. 1, STP-001, STP-003 SP 203.07 A/B/C/D Rev 18 Corrosion test facility operating procedure M-545 Rev. 17, 544 M-532 Rev. 21 SK-6292-E-58 QA surveillance #363 SMUD contract M 13.02-302, M 13.02-295 EM1 A1661, A2175, A2177, A2589, A3534, A3651, A2659. A4786, A5430, A5618, A5651, A5701, R0012. R0048, R0290, R0762, R0968, A-5464, R0328, A3795A-F, A5619 Daily Lab Recorts 6-17-85, 6-3-85, 5-27-85. 5-20-85, 5-13-85, 5-6-85 Technical Specifications Section 1.3, 3.3, 4.5, Procedures AP.103, AP.105, AP.107, AP.108, AP.165, AP.44, AP.91, AP.26, AP.22', i Elant Oceration Procedure i
l A.24 rev. 12, A.51 rev. 31, A.50 rev 18 USAR Section 4.1, 4.2, 4.3, 4.4, 4.5, 5.1, 5.2, 5.5, 6.1, 6.2, 6.3, 6.4, 6.6, 14.1, 14.2, 14.3, 14.4, 14.5, 9.4, 9.5 s
Chemistry & Radiochemistry Manual l
AP.306 III C.2 l
AP.306 III 0.6 10 CFR 50 Calculations l
Z-NSH-M0248 Z-NSH-M0249 Z-NSH-M0250 Z-NSH-M0251 Z-NSH-M0252 Z-NSH-M2077 Z-ZZZ-M2262 Z-EOP-E0064 Z-EQP-E0135 ATTACHMENT 6.1 LIST OF DOCUMENTS REVIEHED Documents System design Basis #5449 - Nuclear Service Cooling Hater System, S08-5448, -5419 SSR - Nuclear Service Cooling Hater - Rev 1, 11/27/86 Casualty Procedure C-33 Vendor Manual "NSCH Pumps" M29.03 1M10 MAP-0009 - Preventive Maintenance Program QAPs (all)
Annunciator Procedures MEL NEAP-4104 - Issuing & Processing of Documents I
DC Procedures Manual, paragraph 9.0 NEAP-4112 - Drawing Change Notice RSAP-0305 - Field Problem Report RSAP-0803 - HR Procedure i
Precursor Analysis EH.ll7A Limitorque Operator Selection Procedures H.ll5/ll6 Spring Pack Test Results ERPT E-0177 QA Serial Work Request Log l
I Hork Request Status log Q28.1 87-136,87-543, 87-575,87-773 l
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ATTACHMENT 6.1 1 I
STATUS OF RIs Attachmtat 6.2 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 SiATUS 104 ACKNOWLEDGED 105 ACKNOWLEDGED l
107 ACKN0HLEDGED 109 ACKN0HLEDGED 110 CLOSED 126 CLOSED 128 ACKNOWLEDGED 132 GPEN 140 OPEN 141 ACKNOWLEDGED
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ATTACHMENT 6.2 1 !
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DETAILED OBSERVATIONS - REOUEST FOR INFORMATION During an evaluation,.all potential concerns are documented on Request for Information sheets (RIs) that are sent to the 4
responsible organization to receive their input concerning the I
potential concern.
RIs are also used to request information that the EASRTP team is having difficulty obtaining.
These RIs are considered drafts throughout the entire evaluation entil 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 l
this section of the report.
k ATTACHMENT 6.3 REQUEST FOR INFORMATION (RI) t RI N0: 104 SYSTEM CODE:
NSCH ISSUE DATE: 08-04-87
SUBJECT:
NSCH COOLANT FLOW IN DECAY HEAT-HEAT EXCHANGERS DEPARTMENT:EUCLEAR ENGINEERING COORDINATOR:
J. ITTNER TEAM LEADER:
M.J. AKINS POTENTIAL CONCERN /00ESTION:
The NSCH flow rate through the shell side of the Decay Heat Removal Cooler may have exceeded the maximum recommended value to prevent tube damage.
Process standard (AP.108) establishes a maximum flow of 3750 gpm in the Dr Heat Removal Cooler, and a system maximum limit of 6600 gpm. A letter was found which states that a similar cooler experienced fatigue due to exces, coolant flow at another B&W plant (B&W to Bechtel, dated 7 Mar 73).
Surveillance data sheets (25 Oct 74 thru 5 Dec 86) show both cooling locpt experience flows between 6000 and 6600 gpm (96% of 119 data sheets). A p' of the surveillance data (TDH V.S. System Flow) indicates a reasonable consistency of data with the furnished pump curves (STP.001-57).
Surveillance flow readings in the past few weeks have shown, using the Controlotron flow meter, that the Reactor Building Emergency Air Coolers :-
receiving ~1050 gpm each.
Using the surveillance data of system flow between 6000 and 6600 gpm, and Reactor Building Emergency Air Cooler flow readings of 1050 gpm each, the resultant flow through the Decay Heat coolers is between 3900 and 4500 gp:
N ATTACHMENT 6.3 i
REQUEST FOR INFORMATION (RI)
RI N0:_l01_
SYSTEM CODE:
NSCH __
ISSUE DATE: 08-04-87
SUBJECT:
NON-CLASS I PIPING TO NSCH RADIATION MONITORS DEPARTHENT:
NUCLEAR ENGINEERING COORDINATOR:
R LAHRENCE TEAM LEADER:
M.J. AKINS POTENTIAL CONCERN /0UESTION:
The radiation monitors associated with both loops of the Nuclear Service Cooling Hater System are attached to the system using non-Class I piping.
The concern is that a failure of this piping can render both loops of this i
Safety Category System inoperable.
Section 9.4.2.1 of the USAR states that the circu1 ting water in cach of -
3 two nuclear service cooling water loops is continuously monitored for radioactivity and an alarm is actuated in the Control Room on high radiat4 i vel.
Further, Section 9.4.3.2 of the USAR states that since non-Class t
components are isolated from C-lass I portions of the NSCH system by Class valves, a failure of the non-Class I subsystems will not compromise the operation of the NSCH system in performing safety-related functions.
A review of the following isometric drawings:
48200-1"-083 Rev. 2, 48201-1"DB3 Rev. 2, 48210-1"-DB3 Rev. O, 48211-1"-0B3 Rev. O, 48220-1"-DB2 Rev. 3, 48221-1"-DB3 Rev. 3; shows that the radiation monitors associated with the Nuclear Service Cooling Hater System are attached to the system '
non-Class I piping.
An examination of Operating Procedure A.24,.Rev. 12 Enclosure 8.1 and 8..
valve lineep show that the Class I valves for isolating the two radiation monitors NSH-079, NSH-080, NSH-081, and NSH-082 are always left in the op?
position.
Since aT1 non-Class I components are assumed to fail in the worst possible state, breaks in all of the lines to and from both radiation monitors can postulated.
This scenario could result in the draining of both NSCH loop the same time and possibly to a low enough level to trip the pumps before breaks could be isolated.
ATTACHMENT 6.3 REQUEST FOR INFORMATION (RI)
RI NO: 107 SYSTEM CODE:
NSCW ISSUE DATE:
03-05
SUBJECT:
NSCH CALCULATIONS FOR FLOW DEPARTHENT:
.NED COORDINATOR:
R LAWRENCE TEAM LEADER:
M.J. AKINS POTENTIAL CONCERN /0UESTION:
Calculations fail to determine correct flow through components in para 11e1 flow configurations.
I Design calculation Z-NSW H0250 failed to' establish proper flow requirement through the one Decay Heat Removal Coolers, and two Reactor Building Emergar.cy Air Coolers Which wore arranged in parallel flow paths.
The unbalanced flows may result.in excessive flows causing tube failure in the Decay Heat Removal Cooler or inadequate flow to meet the heat removal 1
requirements.
Calculation Z-NSW-H0250 assumed a flow rate of 3000 gpm throt the Decay Heat Removal Coolers and 1500 gpm through each of,:
i Reactor Building Emergency Air Cooler without documented i
justification.
Calculation Z-NSW-M0250 determined only the pressure drop requirements with the assumed flow rates of 3000 gpm and 150C gpm separately to meet pump head requirement.
The calculation failed to establish the correct flow rate vai-through each parallel flow paths to verify the assumed values 3000 gpm and 1500 gpm.
The calculation considered only the pressure drop for Reactor Buildint., Emergency Air Cooler A-500C and ignored Reactor Building Emergency Air Cooler A-500A which is in a parallel ansi 20 feet higher in elevation than that of A-500C.
The calculation did not balance all parallel flow paths simultaneous I
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ATTACH"3 T 6.3 l
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REQUEST FOR INFORMATION (RI)
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RI NO: 109 SYSTEM CODE:
NSCH ISSUE DATE:
08-05-87 I
SUBJECT:
HYDRO STATIC TESTING OF THE NSCH DEPARTMENT:
3RTP COORDINATOR:
J. ITTNER TEAM LEADER:
M.J. AKINS POTENTIAL CONCERN /0UESTION:
The NSCH system does not appear to have been pressure tested correctly in accordance with Inservice Inspection Requirements.
ASME Section XI IHD 5000 states:
"The system test pressure shall be at le.
1.10 times the system desian pressure." Hork Request 115720 indicates the hydrostatic test pressure was between 135 to 140 when it should have been above 1 M psig (150 x 1.1).
The only document found in which the system design pressure was identifiec was the operator training manual (Table F, states 150 psig).
1 The component design pressures, stated in various documents, are as follow 1)
The NSCH piping (HE) design pressure is 150 psig at 200 degrees F.
l (Piping design Spec. Sht 7C-25) 2)
The NSCH heat exchanger design pressure is 150 psig.
(NEPM No. 5449) 3)
The NSCH Chemical Mix Tank design pressure is 300 psig.
(NEPH No. 5449) i l
4)
The Decay Heat Removal Cooler design pressure is 160 psig.
(NE.PM No. 5449) 5)
The Reactor Building Emergency Air Cooler design pressure is 200 psi >
(NEPH No. 5449) 6)
The NSCH pump (;:lgn pressure is 110 psig.
7)
The NSCH surge tank design pressure is 100 psig.
3)
The USAR Section 9.4 identifies the system operating pressure as 110 psi 9 ATTACHMENT 6.3 '
REQUEST FOR INFORMATION (RI)
RI NO: 110 SYSTEM CODE:
NSCW ISSUE DATE: 08-06-87
SUBJECT:
GREEN TAG PROCEDURE CONTROL DEPARTMENT: OUALITY ASSURANCE COORDINATOR:
DAVE MALONE TEAM LEADER:
M.J. AKINS 1
1 POTENTIAL CONCERN /0UESTION:
Procedural control of used green tags for controlling QA Class 1, commerc-grade and EQ qualified material, parts and components does not exists, th-providing a way whereby unqualified components could be used in place of qualified components.
SMUD's commitment to 10CFR50 Appendix B, " Identification and Control of Materials, Parts, and Components" may be compromised if the green tagging process, used for controlling all QA Class 1, commercial grade and EQ qualified components does not include voiding, cancellation, or destructi-l of the tags.
While reviewing Work Request package 135203-1, in the Planning Group closure, green acceptance tags for QA Class 1 replacement parts (see Attachments 1 and 2) placed in work package were observed as not bei, voided,~ destroyed, cancelled or otherwise rendered useless.
Nowhere in QA procedures is there mention of proper disposal for gre.
acceptance tag (s) when rtmoved by a QC inspector from ITEM (s) at tir use or installation.
Memorandum J.M.P. 87-16 (see Attachment 3) dated 05-05-87, (an j
uncontrolled document) from QC to all QC inspectors, is the only document dealing with acceptance tag application and removal.
The results of this could be the tagging of Non-Class 1 or EQ components such.
Then, the component.could be emplaced into a safety class system.
This would jeopardize the ability of the system to perform its function.
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This RI was closed because it is being tracked through the Seal Injectior Makeup System RI-124.
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i ATTACHMENT 6.3 l l
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1 REQUEST FOR INFORMATION (RI)
RI N0:_12E_
SYSTEM CODE:
NSCW ISSUE DATE: 08-12-87 SUBJU T:
POSSIBLE VIOLATION OF OAP-17 DEPARTMENT:
MAINTENANCE COORDINATOR:
J. DARKE TEAM LEADER:
M.J. AKINS 1
POTENTIAL CONCERN /00ESTION:
)
QAP-17 was not followed to initiate an NCR when damage, "beyond normal wec and tear," was found during refurbishment of Nuclear Service. Cooling Water pump P-4828.
~When reviewing Work Request 135203-0 continuation sheet, Step 4 A Gauge, 014" deep, 13/16" long and 7/16" wide 12" from outboard face of impeller,
indicated as being found on the shaft.
A gouge in this location could only be caused by the introductions of something through the system; or through the improper handling of the component.
Therefore, this would appear to be a condition beyond normal s and tear, thus requiring an NCR per QAP-17.
)
This was closed based on the response that indicated that the gouge was et by maintenance personnel during refurbishment and that it was evaluated a*
the time by qualified personnel.
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ATTACHMENT 6.3 _
i REQUEST FOR INFORMATION (RI) l l
RI NO: 128 SYSTEM CODE:
NSCW & DHS ISSUE DATE: 08-13-87
SUBJECT:
(NSW) SYSTEM DESIGN CAPACITY DEPARTMENT:
NED COORDINATOR:
RON LAWRENCE TEAM LEADER:
M.J. AKINS POTENTIAL CONCERN /0UESTION:
The Nuclear Service Cooling Water and Decay Heat Removal Systems are not capable of meeting the safety related design requirements as specified in USAR.
The Decay Heat Removal Coolers cannot remove the required heat loac during the post-LOCA mode of operation because the Nuclear Service Coolinc Water inlet temperature is higher than the design value.
The Nuclear Serv Cooling Water Heat Exchanger cannot provide cooling water with the requirs.
temperature since they were designed based on an incorrect Nuclear Service Raw Hater temperature.
The Decay Heat Removal System Design Basis (SDB) specifies a design t transfer of 33.75 x E6 BTU /hr for normal shutdown to reduce reactor coolant from 280*F to 140'F in 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> based on a 96*F NSCW temperature.
(page 28)
The Decay Heat Removal System SDB also specifies an alternate heat ic-condition for post-LOCA of 125 x E6 BTU /hr with a tube side Recirculation Sump Water inlet temperature of 280*F and shell side N:
inlet temperature of 95'F (page 29).
B&W has confirmed that the 125 x E6 BTU /hr is required to meet post-LOCA cooldown when one Reat-Building Eme;gency Air Cooling unit train is lost due to diesel failt-Nuclear Service Cooling Water System AP.108-1 states "the maximum allowable NSCH Heat Exchanger Outlet temperature is 95*F" (Section 1.
The Decay Heat Removal Coolers data sheet (Whitlock Manufacturing Cc.
identifies an inlet cooling water temperature of 95'F and a heat loa 33.75 x E6 BTU /hr, which is the normal shutdown condition.
The Decay Heat Removal coolers would be undersurface with a Nuclear Service Cooling Hater inlet temperature of 125'F instead of 95'F.
i the systems could not be capable of meeting the USAR requirement for 14 hour1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> normal shutdown condition nor is it capable of meeting post cooling requirements.
ATTACHMENT 6.3,
RI 121 (Continued)
)
Calculation Z-NSH-M0251 dated 1969 stated that it would not be possit to supply cooling water at 95'F to the Decay Heat Removal Coolers dur-the accident condition.
With a spray pond temperature of 95'F as specified by USAR instead c:
87'F (NSRH-RI 034) the inlet cooling water temperature to the Decay -
Removal Cooler would be greater than 125'F thus further reducing the system's ability to meet the design requirement.
Based on the above items and RI 025 from the Nuclear Raw Water System, the Nuclear Service Cooling Hater System cannot provide adequate cooling to tr Decay Heat Removal System to meet the safety-related design requirement I
specified in the USAR or SDB.
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ATTACHMENT 6.3 1 )
I
REQUEST FOR INFORMATION (RI)
RI NO: 132 SYSTEM CODE:
NSCH ISSUE DATE: 08-14-87
SUBJECT:
NUCLEAR SERVICE COOLING HATER PUMPS P482A & B DEPARTMENT: MAINTENANCE. MECHANICAL COORDINATOR:
J. DARKE TEAM LEADER:
M.J. AKINS POTENTIAL CONCERN /0UESTION:
CONCERN - The integrity of both Nuclear Service Cooling Hater pumps may ha been compromised due to the work that was performed in accordance with Hot Requests #0135203-0/1 and #0133421-0.
The following events pertain to the work done on the "B" Nuclear Service Cooling Hater Pump.
The pump casing bolts were torqued to values as high as 3 times the manufacturer's specifications and were based only on the stud tensile strength.
The vendor was not contacted to determine casing strength or design characteristics.
The vendor representative indicated concern and stated that since the pump casing is made of cast iron, exceeding manufacturer's torquing specifications can result in stripping the casing threads and possibly a complete failure of the casing.
A casing gasket twice the thickness of that required by the manufacturer'-
specification was installed without contacting the vendor as to possible adverse effects.
The vendor representative stated that increasing the ca.
gasket thickness will result in an elliptical register for the stationary wear rings, which will accelerate wear of the rings, causing premature degradation of the pump's capacity.
A handwritten step was added to repair the mecharsical seal of the pump bu there is no sign-off signature to indicate that the work was performed.
the seal'was not repaired, a failure could occur at any time.
The Installed Parts Equivalency Evaluation Report for the ALL-THREAD, use making the casing studs, was not included in the work package as requirec AP.91 section 7.2.
Therefore, it is not possible to substantiate that tF material used meets ASHE code, section XI requirements for project Class equipment.
Since the scope of the work request for the "A" Nuclear Service Cooling h Pump was identical to that of the "B" pump, the work plan for the "B" was used in planning the work request for the "A" pump.
This resulted in the following conditions which now pertain to the "A" pump.
ATTACHMENT 6.3 _ - _ - _ _.
RI 112 (Continued)
The pump casing fasteners were torqued to a value of three times the manufacturer's specification. A pump casing gasket twice the thickness of that required by the manufacturer's specification was installed during reassembly.
i These conditions bring about the same concerns as those identified in regar j
to the "B" pump.
CONCLUSION:
The work performed in Work Requests #135203 - 0/1 and #133421 may have cause a degradation in the integrity of both Nuclear Service Cooling Water Pumps -
the point that a failure could occur.
NQIE:
This RI is still open because Maintenance is investigating further to determine if the concern is valid.
ATTACHMENT 6.3._. - _ _ _ _ _ - _ _ _ _ _ _ _ -
1 REQUEST FOR INFORMATION (RI)
RI NO: 140 SYSTEM CODE:
NSW ISSUE DATE: 08-17-87
SUBJECT:
VERIFICATION OF LIMITOROUE ACTUATOR THRUST /TOROUE CALCULATIONS DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR:
R. LAWRENCE TEAM LEADER:
M. AKINS I
POTENTIAL CONCERN /0UESTION:
It is not possible to perform a verification of the limit and torque switch settings for MOVs derived from Nuclear Engineering calculation Z-ZZZ-H2262 (draft)
BACKGROUND Engineering calculation Z-ZZZ-H2262 (draft) has been reviewed by Nuclear Engineering and is awaiting issuance by NEDC.
The following deficiencies were identified in association with the calculation.
1.
The calculation references the "proceduri. for the Limitorque operator selection" but does not specifically refer to any numbered or dated procedure from Limitorque.
Therefore, it is impossible to verify the formulae against a specific originating document.
These formulae are the basis for Limitorque MOV torque and limit switch settings to ensur.
that the MOVs will operate adequately in all design conditions.
MOV Nuclear Engineers have stated that a computer code is being generated be included as a revision to the subject calculation package.
This to l
and its associated documentation were not available for review to I
determine if proper reference to original formulae was included.
l 2.
MOV Nuclear Engineers requested the manufacturers of valves which were fitted with M0V operators to provide design input data used in the calculation of torque / thrust values.
The input data received was not attested to by a representative of the valve manufacturer.
Therefore verification of the accuracy of the data is extremely difficult.
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ATTACHMENT 6.3 L
Rl l
REQUEST FOR INFORMATION (RI)
RI NO: 141 SYSTEM CODE:
NSW ISSUE DATE: 08-17-87
SUBJECT:
OC HOLD POINT NOT COMPLETED ON WR 135203 l
I DEPARTMENT:
OUALITY COORDINATOR:
D. MALONE TEAM LEADER:
M. AKINS POTENTIAL CONCERN /0UESTION:
Refurbishment of the NSCW pumps may not have been performed in accordance with the requirements specified in vendor manual.
J BACKGROUND:
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QC Hold Point #5 of HR 135203, Rev. 0 was not signed off prior to HR clos-to assure that the refurbishment was performed in accordance with the venc manual and that qualified parts were installed.
This hold point required QC to verify refurbishment of P-4828 in accordan; with Vendor Manual M29.03-IM10 including the acceptability of replaced parts. Within the comment section of the WR, the QC inspector did refere-attachments to the work request which documented the parts replaced and tr refurbishment verification.
However, several of the parts listed as repl;.
in the work performed section of the HR do not include a QC signature as assurance of verification of parts replaced., and no one signed the hold pr to assure that the entire refurbishment and parts replacement was verifie ATTACHMENT 6.3