ML20237J364
| ML20237J364 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 08/10/1987 |
| From: | Andognini G SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| GCA-87-464, NUDOCS 8708180123 | |
| Download: ML20237J364 (1) | |
Text
$SMUD ..atetivto SACRAMENTO MUNICIPAL UTILITY DISTRICT C P. O. Box 15830, Sacramento CA 95852-1830,NR$6) 4s2-3211 AN ELECTRIC SYSTEM SERVING THE HEAREWON:VLIFORNIA 1981 Al$ II P D 21 I AU('18194 GCA 87-464 U. S. Nuclear Regulatory Commission Attn: J. B. Martin, Administrator Region V Office of Inspection and Enforcement 1450 Maria Lane, Suite 210 Walnut Creek, CA 94596 DOCKET 50-312 RANCHO SEC0 NUCLEAR GENERATING STATION UNIT #1 PRELIMINARY EXPANDED ASRTP INSPECTION REPORTS
Dear Mr. Martin:
Attached for your information are the first six Expanded Augmented Systems Review and Test Program (EASRTP) Inspection Reports. These reports provide the preliminary findings of the EASRTP Inspection team on the following systems:
- Emergency Diesels
- Nuclear Raw Water
- Integrated Control
- Control Room and Technical Support Center HVAC Systems Please note that the affected organizations have not evaluated ror validated the presented findings. All potential corrective actions resulting from the evaluation / validation process will be processed and prioritized through established Rancho Seco procedures.
The District will advise the NRC of any changes to the reports as well as updating your staff on the progress of the entire EASRTP effort.
If you have any questions, please contact Bob Croley of my staff.
Sincerely, I s/byyw G. Carl Andognini Chief Executive Officer, Nuc1 ear 8708100123 870010 PDR ADOCK 05000312 Attachment R PDR 1
cc: G. Kalman, NRC, Bethesda w/atch ) g A. D'Angelo, NRC, Rancho Seco w/atch F. Miraglia, NRC, Bethesda w/o atch gg RANCHO SECO NUCLEAR GENERATING STATION O 1444o Twin Cities Road, Herald CA 95638 9799;(209) 333-2935
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, EXPANDED AUGMENTED SYSTEM REVIEW AND TEST PROGRAM (EXPANDED ASRTP)
EVALUATION
.OF THE NUCLEAR RAW WATER SYSTEM
' DATE:
SUBMITTED BY: ' m M. J. MINS ~ /
TEAM LEADER CONCURRENCE: .> t E- NNw /4 DATE: A- Hr7
/DAVIDHUMENANSKY f EXPANDED ASRTP PROGRAM MANAGER
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s- DATE:
CONCURRENCE: /
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/ B0f CROLEY DIRECTOR, NUCLEAR ECHNICAL SERVICES I
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TABLE OF CONTENTS j P
Pace Number
1.0 INTRODUCTION
3 4
2.0 PURPOSE 3.0 SCOPE 5 4.0 OVERALL RESULTS AND CONCLUSIONS 6 7
5.0 DETAILED OBSERVATIONS - REQUESTS FOR INFORMATION (RI) 11 6.0 ATTACHMENTS 6.1 List of Documents Reviewed
- 6.2 Status of RIs
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EXPANDED' AUGHENTED SYSTEM REVIEW AND TEST PROGRAM 9
EVALUATION OF THE NUCLEAR RAW HATER SYSTEM
1.0 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 [SRTF) and an U analysis of the adequacy of 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 implemented on 33 selected systems and an in-depth review of the engineeri_ng, 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 o and review programs at Rancho Seco.
J-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 ih(iustry standards provided by team members with conLuitants, architect engineer and vendor backgrounds are joined with the specific plant knowledge of SHUD team members.
Each team performs an evaluation on a helected system using the same fundamental evaluation techniques employed by the NRC in the ASRTP inspection. Gystem Status Reports are used as the primary source of leads for the teams. They are augmented with references to availab1_e source and design bases documents as needed. Team synergism and communication is emphasized during the process in order to enhance the evaluation. Each team _ prepares a final report for each completed selected system evaluated. This report is for the Nuclear Raw Hater (NRW) System.
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2.0: PURPOSE,
~ The objectives of the.. Expanded AS'RTP 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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- l. 3.0 SCOPE o
l To accomplish the first objective, the Auxiliary Systems team I evaluated the NRW 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 perforn. 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 1.
The primary source of leads for the team were the problems '
identified in the NRH System Status Report. Various source documents such as the USAR and Technical Specifications and 4 available design bases documents were reviewed as needed to augment i the information needed by the team. j l
The evaluation of the NRW system included a review of pertinent 1 portions of support systems that must be functional in order for the I j
NRW system to meet its design objectives.
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4.0 OVERALL RESULTS AND CONCLUSIONS.
1 LThe 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. Section 5.0' 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. The numbers in' brackets after each individual summary refer to the corresponding RIs in Section 5.0.
The conclusion reached from the review of the Nuclear Raw Hater System is that there is enough inconsistanciesiin the design calculations and the surveillance procedures to raise a concern about the ability of'the system to meet its intended functional requirements.
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OVERALL RESULTS AND CONCLUSIONS (Continued) 4.1 NUCLEAR RAW HATER (NRW) FLOH RATES 4.i.1 A review of the system P&ID H-544, Rev. 20, revealed a lack ,
of any flow measuring instrumentation for the system. The Updated Safety Analysis Report (USAR) and the System Design Basis (SDB) describe the minimum flow rates required through the system and the individual system components.
A review of the Quarterly Surveillance and Inservice Test Procedure (SP.203.07A/B/C/D, Rev. 18) revealed that only the total system flow rate is estimated. A review of inspection data revealed that the method used to determine flow rate yielded data that, in a majority of cases, were beyond the region of pump runout. No documented evidence was able to be produced to show that this was ever questioned. ,
The flow rate in the system is determined by measuring spray header pressure and referring to a flow rate versus pressure graph in the procedure. Nuclear Engineering was not able to provide a source for the curve, nor have calculations been produced to support the curve. A concern exists as to the validity of flow rates in the system throughout the life of the plant and more specifically that individual components such as the Emergency Diesel Generator, (EDG) Jacket Water i Coolers have ever received their minimum required flow.
[RI001] [RIO33] [RIOS6]
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4.1.2 A review of all of the NRW calculations failed to show that adequate flow has been provided through any of the lube oil or room coolers other than through the Nuclear Service Water Cooler.
No calculations could be produced that verify, when the system was designed, adequate flow could be provided to all system components. Individual loop flow rates have been calculated based on the required capacity without considering supply pressure. Because the flow through these loops provide cooling to safety related pumps, lube oil coolers and pump room coolers, improper sizing of the system will reduce the ability of the system to meet the functional support needs of the safety related components. [RIO32]
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1 q l OVERALL RESULTS AND CONCLUSIONS (Continued) 4.1.3 Numerous inconsistencies exist in the calculations which would result in the safety related pump cooling systems being l improperly sized. Similarly, a variety of inlet water temperatures are used in calculations; for example, 87*F for the Nuclear Service Cooling Water Heat Exchanger, 82*F for the EDG Jacket Water Cooler. The Updated Safety Analysis Report (USAR) 1 indicates that 95'F should be used. This agrees with a hi-temp i alarm of 95'F for the spray ponds. [RIO343 [RIO64] j l
4.1.4 The commitment to balance the NRH system flows prior to startup I has been completed. A review of the procedures and observations revealed concerns about the qualifications of the testing personnel, calibration of the equipment and the training for use and care of the instrument. No documents' were able to be produced by System Testing personnel or Training to remove this concern.[RIO20]
4.2 NRW Pond Recirculation Subsystem 4.2.1 The System Status Report (SSR) for the NRH System identified ,
several instances where corrosion, algae and sediment have entered the system and caused plugging of the lube oil coolers and room coolers. No documents could be produced which indicate that a root cause analysis was ever performed to determine the j mechanism (s) which create the problem identified in the SSR. 4
[RIO99]
The surveillance procedures de eloped to monitor cooler plugging do not contain acceptance criteria for determining when plugging is beyond allowable limits. [RIOS6]
4.2.2 'A review of the pond recirculation system revealed a lack of engineering documentation to support the design. Since the only cleanup and chemical control is provided via the recirculation system, no documents could be produced that show that the system was designed to meet the requirements of the USAR. [RIO27]
It was observed that misoperation of the existing system could occur because components are mislabeled and the Spray Pond level switch is improperly sized for controlling the system. [RIO55]
[RIO26] [RIO98) 4.3 Programmatic 4.3.1 Several instances were found where components and systems were released and declared operational prior to closure of the associated Engineering Change Notices (ECN). [RIO21]
4.3.2 Material for safety related components was purchased as commercial grade without the required dedication documentation. ;
[RIO58]
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'5.0' DETAILED OBSERVATIONS - REOUEST 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. l RIs are also used to request information that the EASRTP team is having
- difficulty obtaining.
These RIs are considered drafts throughout the entire evaluation.until they become part of the final report. Responsible organizations can accept the potential concern as valid or they may disagree with the j 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 bet .!r focus the concern. RIs developed during the system evaluation comprise this !
section of the report.
Attachment 2 of the report provides RI status as of this report date. l 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- l valid by the responsible organization.
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-Approximately one week will be provided after the report is issued to provide time for departments to address each RI for validity. A $
revision to Attachment 2 will then be issued to reflect the status of RIs. All RIs not acknowledged at the end of this period will have an i "Open" status. RIs are then transferre'd into the Restart Scope List j tracking system for resolution and corrective action implementation. j l
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.- REQUEST FOR INFORMATION (RI)
RI PK): 001 SYSTEM CODE: NSRH- ISSUE DATE: 7/17/87
SUBJECT:
'ADE00 ATE FLOW TO D/G JACKET'NATER COOLERS
- DEPARTMENT: 'NED ' COORDINATOR: R. LAHRENCE' l
l TEAM MEMBER: M. J. AKINS TEAM LEADER': M . J . A '.
i POTENTIAL CONCERN /00ESTION:
L Please see: attached
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a POTENTIAL CONCERN / QUESTION The existing NSRH Configuration does not provide an accurate means for determining that the Diesel Generator jacket water coolers (D/G J.H.Hx.) are provided with a minimum of 1000 gpm as required by design (ref.
Thermaxchanger specification sheet, see calc. Z-NRH-M0236. file N19.01)
. Flow diagram H-544 rev. 20 does not identify any flow measuring 1 instruments anywhere in the NSRH system. It does provide a !
pressure indicator (local) on the pump discharge line and a I remote pressure indicator (Control Room) on the spray discharge. These do not provide measurement of flow through either the system or the D/G J.H.Hx.
. NSRH Quarterly Surveillance and Inservice Inspection Test Procedure (SP 203.07 A/B/C/D/, Rev. 18, dated 7-25-86) does not
-identify any measurement step for D/G J.H. Hx. flow.
- This procedure does provide a method for estimating total system flow. The flow graph curve provided for this purpose has no identifiable foundation and thus the rerultant flows are questionable.
. Attachment A shows a plot of the system flow versus pressure from the data contained in the surveillance records for the system. As can be seen, most of the data points are beyond pump ;
runout (based on mfgr. pump curve) which is physically impossible.
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. The pump / motor vendors performance curves using pressures and motor ampere readings indicate system flows other than those estimated.
. A walkdown of the system confirmed that no flow measuring devices exist anywhere in the system.
. The 0/G monthly surveillance Test Procedure SP 206.03A, Rev. 16, dated 03-07-86 does not address NSRH flow.
. A visual inspection of the D/G local control panel revealed that a flow indicator for measuring jacket water flow exists on the panel. However, it is not connected. No one could remember why it was not connected or when it was disconnected.
. Discussions with some operators and I&C indicate that the flow indicator has never been connected. The operators felt that it should be.
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1' POTENTIAL' CONCERN / QUESTION -(Continued)
F In summary, there is no way for any operator to determine flow rates through D/G J.W.Hx..at anytime. There appear to be no procedures in place to monitor
- or check flow.' The existing system flow calculating procedure is crude and yields-inaccurate data.
One potential impact ~ stemming from this is: ,
During a loss of offsite power, concurrent with failure of 1 D/G, the potential for losing the second D/G on high Jacket Hater temperature.
caused by insufficient NSRH to the.D/G J.H.Hx. exists..
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L REQUEST FOR INFORMATION (RI)
RI NO: 020 SYSTEM CODE: NSRii ISSUE DATE: 7/22/87
SUBJECT:
-FLOH READINGS ON NSRH TAKEN HITH THE CONTROLOTRON FLOHMETER i DEPARTMENT: STRP COORDINATOR: J. ITTNER __,
TEAM MEMBER: PAUL D. ZELLMER TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /00ESTION:
A concern exits that the liSRH flow balancing and flow rate data will .be of acceptable accuracy. Procedures STP-1031A NSRH Loop.A and B flow balance require flow to be measured. The flow readings for these are being obtained using a Controlotron Ultrasonic Flowmeter. Based upon a review of procedures, calibration data, QA documents and training records, it was found that the plant personnel could not provide:
. procedures for the care and use of the meter
. documents showing approval of the Vendor m.anual
. records of any training of testing personnel on the use', operation and j care of the device.
Field observation has identified that the tracks for-one of the meter points were installed improperly and would yield erroneous flow readings. Also, discussions with personnel installing the flowmeter indicated their.
uncertainty on how to connect and read the instrument.
Because the readings are being used to verify the flow balance of the NSRH i system, which is a Safety-Related System, and because there are no flow devices in the entire NSRH system, it is important that the flow data collected and verified is accurate for baselining the system.
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p REQUEST FOR INFORMATION (RI)-
-RI NO: 021 SYSTEM CODE: NSRH ISSUE DATE: 7/23/87
SUBJECT:
POSSIBLE VIOLATION OF PROCEDURE AP.44 _
DEPARTMENT: NED COORDINATOR: RON LAHRENCE TEAM MEMBER: J. BALDHIN TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /00ESTION:
Systems ~are being turned over to operations while open ECNs exist. This -
could be in violation of procedure AP.44 (Plant Modifications - ECN Implementation, Rev. 11, dated 03-17-87) unless an interim release has been documented. A review of procedure AP.44 states that a system cannot be declared operable until all open ECNs for the system have been closed or an interim release.
. SSR problem #8 states that. Pressure Indicators will be installed across the High Pressure Injection, Makeup and Decay Heat Removal pump lube oil coolers to monitor plugging trends.
. ECN A3795 was generated to respond to SSR #8. It is still open h SDC I
yet the system is in operation and plant personnel were unable to produce documents that confirm that the requirements of AP44 have been satisfied. \
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I REQUEST FOR.INFORMATION (RI)
RI NO:. 025 SYSTEM CODE: NSRH ISSUE DATE: 7/23/87
SUBJECT:
LUBE OIL COOLERS AND PUMP ROOM COOLING HEAT LOADS DEPARTMENT: NED COORDINATOR: RON LAHRENCE TEAM HEMBER: J. BALDHIN TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /0VESTION:
Modifications were'made to the NSRH System without proper review of effect to
' the operating system. Because of this the system may be unable to meet its -
functional objective of supplying cooling to safety related equipment.
. Based on a review of the NSRH P & ID M-544 and a' review of original design calc. no. Z-NRW-M0246, it has been found .that the heat load generated by the emergency pump room cooler, make up pump, High Pressure Injection A & B were not taken into consideration when designing the-NSRH system. It appears that the system was modified without properly analyzing the effect of the change.
Since NSRH is the' ultimate heat sink, the heat loads generated by the ;
aforementioned should have been evaluated prior to implementation of any system changes.
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REQUEST FOR INFORMATION (RI)
RI N0:__02fi SYSTEM CODE: NSRH ISSUE DATE: 7/22/87
SUBJECT:
LOH LEVEL SHITCH-SPRAY POND A (LSL-47001)
DEPARTMENT: SRTP COORDINATOR: J. ITTNER TEAM MEMBER: PAUL D. ZELLMER TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /0VESTION:
NSRH Circulating Pump P-476A is not properly controlled by LSL-47001. P-476A may run without water or may not run sufficiently to keep the pond clean.
The NSRH Circulating Pump P-476A is not controlled adequately. Review of the NSRH SSR, process standards and calibration records indicate:
- Problems #12 and #15 in the NSRH SSR still exist.
. Present leve' switch (LSL-47001) which controls the circulating pump has a range of 0-60" water.
. Process standards AP.107 list the setpoint of LSL-47001 or 0-67" H 2O +
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. ECN A-5619 replaces LSL-47001 and is not planned to be implemented before restart, \
Since the recirculation system is the only means provided for cleaning the spray pond, it is important that it is functioning properly to ensure that debris, sediment, slime, etc. are removal from the system to preclude plugging of components. The system should be capable of the designed function prior to restart.
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REQUEST FOR INFORMATION (RI)
RI NO: 027 '
";(STEM CODE: NSRH
. ISSUE DATE: 7/23/87-
SUBJECT:
NUCLEAR RAW !{ATER SPRAY POND RECIRCULATION AND FILTRATION DEPARTHENT: NED COORDINATOR: RON LAWRENCE
. TEAM MEMBER: D. E. K0ONTZ . TEAM; LEADER: L M .J. AKINS' l'
- POTENTIAL CONCERN /00ESTION:
There is no design bases or calculations to support the present installed ,
recirculate 6 and filteration system eautpment. Further, the system that is '
installed r y not be providing the necessary' flow rate for proper filtration to meet USAR commitments, nor is there any $ instrumentation to determine equipment operability. ,
. USAR Volume IV Section 9.4.2.1 states that each of the ponds has a separate themical feeding and filtering system for continuous clean'.ng of the water.
. Bechtel Design Calculation Z-ilRW-M0157 originally called for a skimmer s system that was' capable of recirculating a volume equal. to the pond ,
width times the pond length ' tin.es a depth of one inch in a '24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time - '
span. Using a 1-1/2 inch pipe and taking suction at one corner of the pc;nd and discharging to the cpposite corner, established the need .for a
, pump with a rated discharge head of 30 feet. The volume requirements' set the pump capacity rating at 20 gpm. The calculation does not
<' address the effects of filtration'. One year after the calculation was complete the pump rating and pipe size were changed for.no bpparent justification. The original project engineer was interviewed and stated -
that"he,could not remember the resson forithe change: . The new figures-Eereforsixinchpipingandal_pumprated'at500gpmat,30feetofhead.
. Te?ephone conversations held with the : filter supplier iridicates that i SMUD ordered a 500 gpm filter and did hot request a filtration system sized for the ponds.
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- At the present rating of 500 gpm recirculation flow would completely
' " recirculate the pond _in approximately four days.
1 . Telephone conversations held with Sacramento County Hater Quality Board Design Engineers indicate that for proper filtration of any body of s
q 's% water, the entire volume must be filtered in each 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, fy
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' POTENil AL CONCERN /0VESTION (Continued).
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. Telephone conversations held with the filter manufacturer confirm this 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> recirculation figure.
. A visual inspection' of the present recirculation / filtration system
.- revealed that there are no pressure, flow, or temperature instruments on
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the system to confirm satisfactory operation.
. Discussions with operators indicate that their only method for determining flow was to listen to the filter for sounds of water flow.
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REQUEST-FOR INFORMATION (RI) t RI N0:__032 SYSTEM CODE: NSRH ISSUE DATE: 7/24/87
SUBJECT:
INCORRECT DESIGN FLONS FOR COMPONENTS IN PARALLEL FLOH PATHS DEPARTMENT: NED COORDINATOR: RON LAHRENCE' TEAM MEMBER: L; JAH TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /00ESTION:
The original design calculations Z-NRH-M0246 and others were incorrect by
-failing to balance all parallel flow paths simultaneously. The design calculations traditionally only verify the main flow and pressure drop to-satisfy the pump requirement, and assume that as long as the main ficw is-adequate, the side stream flow requirement will be met. In this case, the-inadequate flows through the side stream' coolers could result in'the damage of the safety related equipment, and thereby challenge the integrity of the safety feature of the plant. -
. Nurtear service cooling water heat exchanger (1)
Flow - 15,000 gpm 164.62 x E6 BTU /hr q' Source - Z-NRW-H0246
. -Diesel generator cooling water heat exchanger (1)
Flow - 1000 gpm 9.95 x E6 BTU /hr Source - Z-NRW-M0246-
. High pressure injection pump and makeup pump lube oil cooler (2)
Flow - 18 gpm 14.11 x E6 BTV/hr Source - Z-NRW-M0246, B&W 1etter 8/10/70
. Decay heat removal pump bearinc oil cooler (1)
Flow - 10 gpm 1000 BTU /hr Source - X-NRH-M0246, B&W 1etter 6/3/70
. Emergency pump room air coolers (3)
Flow - 35 gpm 170,000 BTU /hr Source - Vendor Data Sheet, Marlo Coil works, 4.12.71.
. Reactor building spray pump bearing oil cooler (1)
Flow - 10 gpm (assumed)
Source NRN-M0246 i
POTENTIAL CONCERN /00ESTION (Continued):
All flows through the components and coolers are arranged in parallel flow path. The only design calculation (A-NRH-H0246) available prior to the restart program is the main flow through the nuclear service cooling water heat exchanger, where the pressure drop and flow were calculated to meet the nuclear service raw water pump head requirement. The calculation also include pressure drop through the diesel generator cooling water heat exchanger. All parallel flow paths should be balanced simultaneously to determine flow rate through each path for pump head requirement.
. Survey of all NRH design calculations fails to show any calculation to demonstrate adequate flow through the components and coolers other than the nuclear service cooling water heat exchanger.
. Recent calculation Z-NRH-H2260 balancing nuclear service cooling water heat exchanger, decay heat removal pump gearing oil cooler, and emergency pump room air cooler in a parallel flow paths, shows flows
, less than the design value of 10 gpm for the decay heat removal pump bearing oil cooler, and 35 gpm for the emergency pump room air cooler.
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. Flow tests dated June 24, 1987, indicated all parallel side stream flows through the coolers were below the design flows as shown above, except the diesel generator cooling water heat exchanger.
. Flow tests also showed that there were inadequate flows through the side streams even with the valve at the main flow through the nuclear service l
cooling water heat exchanger completely closed.
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In summary, the original design calculations Z-NRW-H0246 and others were incorrect in failing to balance all parallel flow paths simultaneously.
Instead, only the main flow of 15,000 gpm through the nuclear service cooling water heat exchanger was verified to meet the pressure drop requirement of the pump. It appears the calculation assumes that there will always be sufficient flow through the side stream which is approximately 100 gpm by throttling the valve at its main flow. The assumption is true only when the side stream flow resistances are relatively low.
REQUEST FOR INFORMATION (RI)
RI NO: 033 SYSTEM CODE: NSRW ISSUE DATE: 7/23/87
SUBJECT:
SP.203.07 A/B/C/D DEPARTMENT: SRTP COORDINATOR: JOHN ITTNER TEAM HEMBER: D. E. KOONTZ TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /00ESTIDS:
Unverified data:is being used to determine system flow rates in a safety related system. 0n October 6, 1974'the "A" side nuclear service cooling water surveillance SP.203.07C, was performed, and the results turned over to the plant mechanical engineer for evaluation. Seventeen days later the system engineer determined that the cooling water pump had not met the required flow rate. On November.5, 1974 an abnormal occurrence report was filed with the Atomic Energy Commission. In the report SMUD decided that the surveillance procedure would be revised to include " acceptance criteria" so that the operator who records the data wouid be able to evaluate the results. On November 18, 1974, Mr. Rodriguez directed the Plant Mechanical.
Engineer, to handle revising the procedure. On December 4, 1974 the procedure was revised to include Enclosure 6.7 (Attachment #2), a flow versus {
spray header pressure graph. The plant personnel were not able to provide verified data supporting the generation of the graph. This graph is still being used to determine total flow. \
Since this graph is the only means for determining system flow it is important that it be as accurate as possible and represent all possible system conditions which may exist at the time of surveillance.
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REQUEST FOR INFORMATION (RI)
RI NO: . 055 SYSTEM CODE: NSRH ISSUE DATE: 7/27/87
SUBJECT:
PHYSICAL IDENTIFICATION OF MOTORS. PUMPS. AND INSTRUMENTATION {
l DEPARTMENT: SRTP COORDINATOR: JOHN ITTNER TEAM MEMBER: R. L. V0H ESCHEN TEAM LEADER: M.J. AKINS POTENTIAL CONCEBN/0UESTION:
Equipment and components necessary for system operations are not properly tagged or labeled. Since f.ield mounted equipment (motors, pumps, instruments, etc.) are not individually identified, the validity of identification for operation or repair of said equipment on approved documents (open and those closed out) are in question. The mis-operation, replacement, or repair of designated equipment for any field action is then in question. The mis-operation, exchange in shop repair, and calibration /modblation of specific equipment could affect the plant ,
functionality and create inaccuracies in the Master Equipment List.
- The existing NSRH system motors, pumps, and instrumentation, etc., lack a physical identification of equipment. In a walkdown, two pumps were s
found having been identified by hand painting the designation. ) l
- Of the two hand painted pump designations, one is found badly faded, and the second was found unreadable due to the peeling of the base paint.
. The NSRH Circulating Water Pumps P476A and P;4768 have the same identification: Westinghouse, 7.5 hp, model number 71D15449, SN 7105.
Both plates read the same.
- In a supplemental walkdown, equipment was randomly selected without l regard to system or physical location, and no designation on the small equipinent was found. An exception was found, in that the very large pieces of equipment were found to have been stencilled with a designation.
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L REQUEST FOR INFORMATION (RI)
RI N0:....064 SYSTEM CODE: NSRH ISSUE DATE: 7/28/87
SUBJECT:
PUMP ROOM COOLER ANALYSIS AT REDUCED NRH FLOH DEPARTMENT: NED COORDINATOR: RON LAHRENCE TEAM MEMBER: L. JAH TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /00ESTION:
Verified and' approved calculations are being used to support restart which contain fundamental errors. Z-HVS-M2258 was performed to justify 16 gpm instead of 35 gpm and an inlet temperature of 95'F for nuclear' service raw water.
. It appears that there is no heat balance between the water side and the air side of the cooler, as evidenced by the results of the calculation page 12 and page 9.
. The water side temperature increases from,95'F inlet to 134*F outlet, to give a total heat removal capacity of the cooler 312,000 BTU /hr (page 9), which contradicts water side heat load of 312,000 BTU /hr.
. It should be recognized that 176,273 BTU /hr is a maximum design heat load, not the heat removal capacity of\an existing cooler with a given surface, nor should it be assumed so.
. A room air temperature of 130*F under these conditions is not acceptable, as it exceeds the normal harsh environmental temperature of 108'F (Engineering Report ERPT-E-187).
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REQUEST FOR INFORMATION (RI)
- RI N0: 063 SYSTEM CODE: NSRH ISSUE DATE: 7/28/87-
SUBJECT:
DESIGN GUIDE REFERENCE TO NO EXISTENT DESIGN CRITERIA DEPARTHENT: NED COORDINATOR: RON LA'WRENCEL TEAM HEMBER: R.-L. VON ESCHEN TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /0UESTION:
'The procedures used for determining Environmental Qualifications via the Design Guides refers to another criteria document that:does not' exist. ,
< .- Design Guide 5204.51-(approved 09-'23-85) " Limitations on use of electrical equipment and materials in a Nuclear Containment," paragraph 5.3. and 6.1, Reference Design Criteria 5101.4 " General Design Criteria -
Environmental,.as the guide for material qualification. Design Criteria 5101.4 is said by Nuclear Engineering. to not exist.
- Engineering Report #ERPT-E0187 " User Manual for the Rancho Seco Environmental Equipment Qualification Program, approved in Engineering
. dated March 27, 1986, was. offered as the alternate to the Design 4 Criteria 5101.4 by the Environmental Qualification Engineer in the Electrical Engineering group.
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1 REQUEST FOR INFORMATION (RI)
RI NO:__Q31 SYSTEM CODE: NSRH ISSUE DATE: 7/28/87
SUBJECT:
HCRS-4969 DISPOSITION AND CORRECTIVE ACTION DEPARTMENT: SRTP COORDINATOR: JOHN ITTNER TEAM MEMBER: J0EL HESTVOLD TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /00ESTION:
An NCR was closed out without all the requirements being completed. The dispositica of NCR 4969 requires revision of SP.203.07 to include the recommendations identified in -Z-NRH-M1735. A review of the SP and the calculation indicate that the SP has been changed to require measurement of the pressure drop across the NSCH Heat Exchanger tubes and has included the '
acceptance criteria for the NSCH Heat Exchanger tube delta P and the NSRH spray nozzle delta P. However, the calcul7 tion also addresses that the cumulative pressure drop across the heat . changer tubes and spray nozzles should not exceed 5 psig. This cumulativt pressure drop has not been addressed in SP.203.07. Case I in the conclusions section of the calculation providas an example of how individually each component (tubes and nozzles) may be within the acceptance criteria, but the cumulative pressure drop exceeds the acceptance criteria and, therefore, per the calculation the t'
" system should be shut down and investigated." This case would not have been identified using the existing SP.
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REQUEST FOR INFORMATION (RI)
RI NO: 057 SYSTEM CODE: NSRH ISSUE DATE: 7/28/87
SUBJECT:
NONCONFORMANCES IDENTIFIED ON ODRs DEPARTMENT: LICENSING COORDINATOR: JERRY DELEZINSKI
. TEAM HEMBER: J0EL HESTVOLD TEAM LEADER: M.J. AKINS l POTENTIAL CONCERN /00ESTION:
NCR's are not to be written against ODRs thus violating QAP17. .0f the 20 NRH ODRs issued during 1986 and 1987, eight required NCRs to be written per the requirements of QAPl7. However, NCRs were identified for only four of those eight ODRs. The other four (86-263,86-338, 86-366, and 87-753) do not appear to have had NCRs written against them even though nonconforming conditions existed.
A review of the closure process for these four ODRs indicates that some control exists to assure that the'necessary corrective action is completed.
However, the level of review for both the discrepancy description and disposition do not appear as extensive as is required for an NCR. In fact, it could not be determined who reviews ODRs to assure that the response adequately addressed the original discrepancy.
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The existing procedure, AP.22, Rev.12, requires the ODR originator to determine if an NCR should be initiated. Apparently AP.22 is being revised to raquire Licensing review of discrepancies for NCR considerations but this revision has not yet been issued.
During this ODR review it was also identified that the ODRs were not being stored properly as QA records in violation of RSAP-0601. The ODRs were not filed in fire-proof cabinets prior to turnover to RIC.
REQUEST-FOR INFORMATION (RI)
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RI'NO: 058 SYSTEM CODE: NSRH ISSUE DATE: 7/28/87 l
SUBJECT:
PROCUREMENT OF LUBE OIL COOLERS FOR ECN A-3795
[ i DEPARTMENT: NED COORDINATOR: RON LAHRENCE TEAM MEMBER: J0EL..WESTVOLD TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /00ESTION:
Unqualified materials which may not be able to meet their intended functional requirements are being used in safety related systems. Per RIDR, QA #1236, the lube oil coolers were purchased as Commercial Grade parts used in QA Class 1 safety applications is required by 10CFR21. Standard industry practice defines dedication as the identification of attributes critical for ;
the component to perform its safety function and the determination / 4 implementation of methods to assure that these attributes are verified as acceptable, i.e., visual inspection, testing.
Plant personnel have not been able to produce' documentation to substantiate dedication of these coolers. The DBR briefly indicates that the difference ;
in heat removal between the old and new' coolers would not affect design. A i
-letter from the cooler supplier states that the coolers are dimensionally
. interchangeable. However, this documentation does not fully address the cooler dedication. \ '
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8 REQUEST FOR INFORMATION (RI)
RI NO: _01B SYSTEM CODE: NRH ISSUE DATE: 7/31/87
SUBJECT:
MINIMUM SPRAY POND LEVELS DEPARTMENT: NED COORDINATOR: RON LAWRENCE TEAM MEMBER: D. E. KOONTZ TEAM LEADER: M.J. AKINS POTENTIAL CONCERN /0VESTION:
The low level alarm setpoint for the Nuclear Raw Hater Spray Ponds is below the minimum level required to met the seven day, post LOCA, criteria.
Process standards states that the level switch which causes the annunciator to sound in the Control Room is set for 5'7" + 1"/-3". This means the alarm could occur anywhere between 5'4" and 5'8". The purpose of this alarm is to indicate that the volume in the pond is at the minimum level required to meet the System Design Bases for a 7 day period of operation, post LOCA, without makeup water. The alarm setpoint was established by Z-NRW-M0243. This document states that the minimum volutae required to meet the 7 day criteria is 2.69 x million gal which calculates out to S.6' or 5'7.2". The minimum volume necessary to meet the 7 day criteria is calculated Z NRW-M0245. It states that the minimum volume required is 2.745 x million gal which calculates out to about 5.72' or 5'8.64".
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In summary, at present the spray pond levels can go below the minimum volume required to meet the 7 day, post LOCA scenario.
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3' REQUEST FOR INFORMATION (RI)
RI NO: _0.93 SYSTEM CODE: NRW ISSUE DATE: 7/31/87
SUBJECT:
ROOT CAUSE ANALYSIS OF LUBE OIL COOLERS PLUGGING DEPARTMENT: NED COORDINATOR: -RON LAWRENCE TEAM MEMBER: M. AKINS TEAM LEADER: M. AKINS POTENTIAL CONCERN /0UESTION:
A root cause analysis has not been performed to determine the cause of the plugging of lube oil coolers for safety related components. 2
. SSR Problem 6 addresses bearing temperatures trend analyses-for the High Pressure Injection Pump, Makeup Pumps, Decay Heat Pumps and Reactor Building spray pumps. This problem was caused by plugging of lube oil coolers thus, reducing cooling water flow to the coolers.
The problem resolution is to implement a treNing program using- 'I temperature readings on bearing cooling water inlet, not outlet. !
.. SSR Problem 7 addresses monitoring heat exchanger performance on NSRW cooled heat exchangers. This was recommended because of heat exchangers plugging.
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The resolution for this problem is to modify the surveillance procedure.
to obtain data to trend heat exchangers performance via pressure drop, not flow reduction. However, no acceptance criteria is addressed in the new surveillance procedure.
. SSR Problem 8 addresses a program to measure tube side pressure drops on exchangers with tubeside NSRH cooling. The purpose of this is to measure partial flow blockage.
The resolution provided for this is to install test connections and revise surveillance procedure to acquire data necessary to trend heat exchanger pressure drop, not flow rate reduction. Again, no acceptance criterion has been established.
Each of these SSR problems address the same thing, plugging. No 1 documentation could be produced where a root cause analysis was done or is being performed to determine the cause of the plugging. The solutions to the SSR problems do not resolve the problems, nor do they address the fluid dynamics of the system which indicate that the present flow rate may not be adequate with no plugging or fouling. They only address provisions to monitor the plugging / fouling happen. This does not ensure that the system j
will meet or be ready to meet its intended function during an unexpected incident. {
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J 6.0 UACHMENTS AT s
6.1 List of Reviewed Documents 6.2 Status of RIs I
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.________________D
.j E LIST OF REVIEWED DOCUMENTS o
- STP.1031 B, 1084 . .
ECN A3795,'A3795A,B,C.D.E,F, R0328 A4905..A5619, A5651 LER 81-16, 87 ODR'86-263,86-386, 86-388,86-366, 86-276,87-282, 86-167,86-396, 86-386,86-394, 86-457,87-220, 87-241,87-404, 87-506,87-560, 87-638,87-746,
.87-753,87-760, 87-764
.A0R 50-312 74-7, 50-312-74-9 E-480A&B Vendor Manual & Bid Specs A-529A,B,C,D,E Vendor Manual & Bid Specs P-472A&B Vendor Manual & Bid Specs P-476A&B Vendor Manual'& Bid Specs F-475A&B Vendor Manual & Bid Specs P-471A&B Vendor Manual-& Bid enecs
.P-238A&B Vendor Manual P-261A&B Vendor Manual P-236 Vendor Manual P-291 Vendor Manual Controlotron Vendor Manual Controlotron QA Manual Chemical / Radiation Logs 01-01-84 through 12-30-84 P0 RS-37743, RS-93501 Master Equipment List (MEL)
RJR Letters74-440, 74-445 .j MSRC Heeting Agenda 12-05-74 TS/600/33 Calibration Records for LSL-47001 USAR .
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Technical Specifications :
Process Standards System Operating Procedures System Design Bases Annunciator Procedures Casualty Procedures Chemistry Manual Station Manual 0A Manual System Status Reports Bechtel Design Calculations Z-NRW-M0236 Bechtel Design Calculations Z-NRH-M0246 Bechtel Design Calculations Z-NRW-M0237 ,
Bechtel Design Calculations Z-NRH-H0238 l Bechtel Design Calculations Z-NRH-M0157 Bechtel Design Calculations Z-NRH-M0243 Bechtel Design Calculations Z-NRH-M0245 Bechtel Design Calculations Z-NRW-M1729 Bechtel Design Calculations Z-NRH-M1735 Bechtel Design Calculations Z-HVS-H2258 Bechtel Design Calculations Z-NRW-M2260 ATTACHMENT 1 l
l LIST OF REVIEHED DOCUMENTS (Continued)
G P& ids M-544, Rev. 20 AP.22, 33, 44, 46, 306 System Training Manual j System Lesson Plan SP.203.07A/B/C/D, 206.03A HR 69886, 76082, 62926, 63190, 76081, 98766, 77882, 69884, 76090, 98769, 69885, 69882, 76094, 78839, 78771, 69883, 98768, 76093, 7784, 60503, 83834, 98770, 76089, 54052, 54047, 64809, 65027, 60501, 49753, 64807, 65028, 60503, 49754, 56338, 64808, 71209, 49752, 54050, 54051, 60016, ,
63052, 77929, 002340, 014103, 019929, 54135, 66873, 68902, 014115 -
014148, 71440, 77910, 92880, 98730, 98766, 98768, 98770, 99956, 104680, 105120, 107435, 107824, 110755, 111359, 111840, 111843, 113174, 113175, 113176, 113981, 113982, 113983, 114587, 114611, 114619, 114900, 115333, 115334, 115341, 115346, 115719, 115761, 115760, 115795, 115809, 115864, 116232, 116662, 117254, 117255, 117256, 117257, 117258, 118036, 118037, l 118038, 121770 NCR 1589, 2039, 6264, 4780, 2898, 6664, 3634, 4489, 3884, 4303, S2522 through S2527, 53178, S4565, S4726, S4727, 4913. S4969, S5023, 5242, 5417, S5425, S-5645, S5464, 5499, 5584, S5645, S6103 Engineering Report #ERPT-E0187 GVC-87-056 {
NUREG 800, Rev. 2 i Precursor Review Task Final Report RIDR QA #1236, #3590 QA Audit 0-494, 87-10, 87-21, 87-23, 87-24 M30.01-5 Approved Suppliers List \
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ATTACHMENT 1 1
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STATUS OF RIs '
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4 RI NUMBER STATUS RSL NUMBER
-001 0/ACK RSL-RI-001
.020' 'O 021 .0
.025 0/ACK- RSL-RI-025 026 0/ACK RSL-RI-026 s- . 027 0-032' '0/ACK- ' RSL-RI-032-033 0 034 0/ACK' RSL-RI-034
-055 0 056 0 057 0' ..
058 ~ 0/ACK RSL-RI-058 063 - 0/ACK RSL-RI-063-064 0/ACK.- ' RSL-RI-064 098 0 099 0 ,
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i ATTACHMENT 2
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F EXPANDED AUGMENTED SYSTEM REVIEH AND TEST PROGRAM (EXPANDED ASRTP) s x
EVALUATION OF THE l'
CONTROL ROOM / TECHNICAL SUPPORT CENTER ESSENTIAL HVAC SYSTEM SUBMITTED BY DATE: E 3-87
[/tFREN R. E$PERpA TEAM LEADER g
CONCURRENCE: &8N % mc DATE: R-FJ7 D HUMENANSKY
/ AVID / .c
/ EXPANDED ASRTP PROGRAM MANAGER CONCURRENCE: / \ DATE:
' /
/ B0B(CROLEY DIRECTOR, NUCLEAR T HNICAL SERVICES PT Giii i if o TST^ _ . _ _ _ _ _ _ __-_____ __ a
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- TABLE OF CONTENTS' s
Paae Numb'er;
- l'0? INTRODUCTION
. 3 2.0 PURPOSE 4 f-3.0 SCOPE. -5.
4.0 LOVERALL RESULTS AND CONCLUSIONS .6 5.0 DETAILED OBSERVATIONS - REQUEST FOR INFORMATION 9 6.0- ATTACHMENTS 27
, 6.1 List' of Documents Reviewed 28 6.2 Status of RIs 31
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f EXPANDED AUGMENTED SYSTEM REVIEH AND TEST PROGRAM EVALUATION OF THE CR/TSC ESSENTIAL HVAC SYSTEM i.0 INTRODUCTIQH 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 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 implemented on 33 selected systems and an in-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 i members with consultants, architect engineer and vendor backgrounds are joined with the specific plant knowledge of SMUD team members.
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Each ten performs an evaluation on a selected system using the same l 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 enhance the evaluation. Each team prepares a final report for each completed selected system evaluated. This report is for the Control Room / Technical Support Center Essential (CR/TSC) HVAC i System.
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t-k 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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5 k . S '. 0 SCOPE To accomplish the first objective, the HVAC team evaluated the CR/TSC ESSENTIAL HVAC 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 HVAC team reviewed the programs as implemented for the CR/TSC ESSENTIAL HVAC system in the following functional areas: !
- 1. Systems Design and Change Control
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- 2. Maintenance
- 3. Operations and Training
- 4. Surveillance and Inservice restin3
- 5. Quality Assurance
- 6. Engineering Programs The HVAC team reviewed a number of documents in preparation for and during the Expanded ASRTP evaluation. This list of documents is found in Attachment 1, Section 6.0.
The primary source of leads for the team were the problems identified in the CR/TSC ESSENTIAL HVAC 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 CR/TSC ESSENTIAL HVAC system included a review-of pertinent portions of support systems that must be functional in order for the CR/TSC ESSENTIAL HVAC system to meet its design objectives.
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4.0 OVERALL RESULTS AND CONCLUSIONS I(
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. Section 5.0 provides detailed findings by providing the Request for Information (RI) forms that are used by the Expanded ASRTP teams to identify '
i potential concerns during the evaluation. The numbers in brackets after each individual summary refer to the corresponding RIs in Section 5.0.
4.1 The Control Room /TSC Essential Filtration System 4.1.1 There is a potential for the CR/TSC Essential Filtration System HEPA filter bank to become overloaded during a high radiation event. In the high radiation mode 50% of outside air (contaminated) is mixed with 50% clean return air upstream of the moisture separator to capture particulate (such as dust) which could overload HEPA filter bank.
In the Rancho Seco CR/TSC Essential Filtration System, the filtration units SF-A-7A and SF-A-78 each have a combination of moisture separator /prefilter in the housing. Performance documentation of the fiberglass pads in these moisture separator sections is inadequate. The installed pads may $
not meet the prefilter efficiency requirements of ANSI N509-1980. Prefilters shall have an average dust-spot efficiency of not less than 45%. \The efficiency of the moisture separator in the CR/TSC essential filtration unit is less than 45%. A review of the system also indicated that the present instrumentation and surveillance may not detect dirty moisture separator /prefilters. Additionally, there are no signals or alarms that record or inform Control Room operators of pertinent pressure drops and flow rates.
As a result, airborne dust could buildup in the face of the HEPA filter bank causing excessive pressure drop and reduced fan capacity. Reduced fan capacity could cause CR/TSC space pressure to fall below 0.125 inches water gauge higher than the CR/TSC boundary (design). (RI #006) (RI #045) 4.1.2 A review of the system indicates that during a high radiation event, the TSC may not be maintained at a positive pressure of at least 0.125 inches water gauges relative to the CR/TSC boundary as required by the design bases. During a high radiation event and toxic gas event, the CR/TSC Essential HVAC system is actuated and the TSC Normal Air Handler (AH-A-2) continues to operate as required, Infiltration of unfiltered air through isolation dampers HV-54717 and HV-54718 can occur which makes static pressure on the Auxiliary Building side greater than the TSC side.
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OVERALL RESULTS AND CONCLUSIONS (Continued)
As a result, continued inleakage may exceed the amounts j assumed for dose and chlorine concentration and could cause j the area to become-inhabitable. (RI #065) j 4.2 The Control Room /TSC Essential HVAC
-l 4.2.1 The System Status Report Problem 39 identified the CR/TSC i Essential HVAC controls and operating status indication in. i the Control Room as being inadequate. ]
A review of the CR/TSC HVAC system indicates that:
- 1. There is no local or Control Room-annunciation to inform the operator to manually reset the condensing unit relay switch during power interruption. As a result, the condensing unit will not autostart upon receipt of Hi Temperature Hi Radiation, or Toxic Gas signal.
- 2. There is no annunciation for compressor crankcase, ,
heater failure (such as blown fuse). The compressor l unit will automatically start upon receipt of Hi i Temperature, Hi Radiation or Toxic Gas signal.
Operating the compressor without reaching the required crankcase temperature could cause a low oil pressure f and excessive bearing wear due to lubrication with oil which contains refrigerant. \
Technical Specifications 3.13 and 4.10 requires the CR/TSC Essential HVAC system be operable during normal plant operations or the plant must be shutdown within the specified time limits. Consequently, the CR/TSC Essential HVAC system could.be inoperable and in violation of Technical Specifications without plant personnel being aware i of the condition. (RI #042) 4.2.2 The CR/TSC Essential HVAC System Status Report, Problem 9, states, "The refrigeration system arrangement does not accommodate convenient or efficient maintenance of the f system." ]
Few provisions have been made for the removal and l replacement of heavy HVAC sections on the Auxiliary Building roof.
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l0VERALL RESULTS AND CONCLUSIONS' (Continued)
At present there are no permanent cranes, hoists or monorails on the Auxiliary Building roof to facilitate the removal of essential compressors, motors, fans or coil. The.
absence of rigging equipment will increase the duration of major repair and could cause plant shutdown. Rancho Seco Technical Specifications, Section 3.13.3, requires the plant ,
to go to hot standby if one CR/TSC Essential HVAC Train is
'noperable for more than 7 days ; or both CR/TSC Essential HVAC Trains are inoperable for more than 3.5 days. l A draft Maintenance Procedure (H.160) was written for the removal and reinstallation of compressors and motors. This ;
draft procedure is inadequate because of the ability of the l Bantam Crane to pick the load in the limited available i space. If the plant had to restart, renting and setting a i larger _ crane in order to complete needed maintenance would increase repair time by days. This could also necessitate a ,
plant shutdown by Technical Specifications. (RI #040) 4
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( 5.0 Qf~Lb1 LED OBSERVATIONS --RE00EST FOR INFORMATION
- During an evaluation, all. potential concerns are documented on c . 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 drafts throughout the entire evaluation ,
until they become part of the final report. Responsible i 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 i members to better focus the concern. RIs deve?oped during the system evaluation comprise this section of the> report.
Attachment 2 of the report 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.
Approximately one week will' be provided after the report is issued- .
to provide time for departments to address each RI for validity. A 2 revision to Attachment 2 will then be issued to reflect the status h of RIs. All RIs not acknowledged at the end of this period will !
have an "Open" status. RIs are then transferred into the Restart Scope List tracking system for resolution and corrective action implementation.
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REQUEST FOR INFORMATION (RI)
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RI NO: 006 SYSTEM CODE: CR/TSC HVS ISSUE DATE: 07-21-87
SUBJECT:
MCR/TSC ESSENTIAL FILTRATION UNIT PERFORMANCE DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR: RON LAHRENCE TEAM MEMBER: J. S. GRILL TEAM LEADER: EFREN ESPERANZA EQIENTIAL CONCERN /0VESTION:
The documentation of performance of the fiberglass pads in the moisture separator sections of filtration units SF-A-7A and SF-A-78 is inadequate. The fiberglass pads may not meet the prefilter effi_ciency requirements of ANSI-N509-1980.
SMUD Design Criteria 5108.10,'Section 5.4.1, states that the safety-related air filtration systems shall be designed in accordance-with the requirements of Reg. Guide 1.52, Rev. 2 ERDA 76.21, ANSI N509 and IEEE 323.
Reg. Guide 1.52, Section B, states "prefilters remove the larger particles and prevent excessive loading of the HEPA filters; to some extent demisters may also perform this function." ANSI N509-1980, y
_ Section.4.1, _ states "prefilters are required in units where design inlet particulate concentrations and particle size are such that the HEPA filter may be rendered ineffective." Section 5.3 states that prefilters shall have an average atmospheric dust-spot efficiency of not less than 1
451.
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The_ efficiency of moisture separators in capturing particulate is
.usually below 45% by weight. .This means that most of the particulate
- (such as dust) entering the essential filtration units would go through the moisture separators and collect on the face of the first HEPA bank.
In the high radiation operation mode, 1,760 cfm of outdoor (contaminated) air is mixed with 1,760 cfm of clean return air upstream of the moisture separator.
During a long term radioactive release and continuous filtration, airborne dust may build up on the face of the first HEPA filter bank causing excessive pressure drop and reduced fan capacity. Likewise, high winds in conjunction with a radioactive release could overload the HEPA filter bank with dust. Reduced fan capacity could cause the MCR/TSC '
space pressure to fall below 1/8 inch w.g. (design).
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-g, RI-00 (Continued)
The in-place DOP test-for HEPA banks utilizes 0.3 micrometer size particles (smoke). This test would not determine the effect af heavy L
dust loading. ' Likewise, IEEE' 323 qualification tests normally do not L, consider the effects of particulate.
In summary, the dust removal efficiency of the moisture separh.to s should be determined and. .if less than 45%, additional analysis should be
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performed.
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- REQUEST FOR INFORMATION (RI)'
1RI N0:- -009 SYSTEM CODE: CR/TSC HVS ISSUE DATE: 07-22-87 E ..s e
.e
SUBJECT:
_ LABELS ON 'A' CONTROL ROOM /TSC ESSENTIAL HVAC EOUIPMENT .
1 jh 'DEPARTHENTE OPE 8ATIONS C00RDINATOR: R. MACIAS TEAM MEMBER; M.L,MCMD TEAM LEADER: EFREN ESPERANL,A .
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s y: 3, ' 'i EQTENTIdlCONCEkN/00ESTh$: ,
Equipment labels were found not to be in' conformance with Administrative
".. . . Procedure AP.23.15,in several cases during a partial system walkdown on V
the ' A' Train Essentidl Control Room /TSC HVAC.
+ Martual isolation valves HVS-007, HVS-009, HVS-Oll, HVS-017 and HVS-0?9,(P&ID M504 T-4) all lack valve identification tags.
. Breaker 2A32? label. indicates the load on ttat breaker but does not
, $s indicait it <is breaker "2A323".t Breaker ID is written on the door wi th " magics marQr?"
. Breaker 1A324 labei does not list HV-54727 as one of its loads as-1i indicated on E-N7 Sh. 27.
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t, REQUEST FOR INFORMATION (RI)
RI NO: 013 SYSTEM CODE: CR/TSC HVS ISSUE DATE: 07-28-87
SUBJECT:
MATERIALS USED TO SEAL THE CR/TSC ESSENTIAL HVAC SYSTEM PRESSURE ENVELOPE DEPARTHENT: _SXSTEM ENGINEERS C0ORDINATOR: JOHN ITTNER TEAM MEMBER: B0B THOMAS TEAM LEADER: EFREN ESPERANZA POTENTIAL CONCERN /00ESTION:
The Hardcast tape used.to seal the essential air. handling units and the Dow Corning 732 ' sealant used to seal the essential duct joints has not been documented as being qualified for long term exposure to the environment.
ANSI /ASME N509-1980; Nuclear Power Plant Air Cleaning Units and !
Components, Section 4.12 states: j "All ESF housings shall be welded. Transverse joints of ESF ducts may be welded or made with gasketed flanges."
Regulatory Guide 1.52; Design, Testing, and Maintenance Criteria for Post n Accident Engineered-Safety-Features Atmospheric Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants, Section Sc states: \
l "The use of silicone sealants or any other temporary patching i material on filters, housing, mounting frames, or ducts should not be allowed." i However, Dow Corning 732 sealant was used during the installation of the ducts to sul the flanged and gasketed joints. Also, during the startup testing of the Essential HVAC System, it was revealed that the air i har.dling units could not meet the leakage criteria of ANSI N509-1980 due i to inadequate design and fabrication of the unit housing. "Hardcast" l tape was used to seal all of the joints of the housing based or, assurance i of qualification from the manufacturer of the tape. Upon evaluation of l the qualification information available, it was determined that there was ,
insufficient data available to establish a qualified life of the tape. !
The inadequate qualification of Hardcast Tape and Dow Corning 732 Sealant has been documented by Problem 33 of the CR/TSC Essential HVAC System l Status Report and NCR S-6576 respectively. Problem 33 states the "Hardcast" tape used to seal the Essential Air Handling Units against air infiltration has not been documented as being qualified for long term.
exposure to expected environmental conditions.
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.y RI-013 (Continued) 7 1.l Procedure Rt-HVS-Oll, Refueling" Internal Inspection sf Hardcast Tape on Control Room / Technical Support Center (CR/TSC) Heating, Ventilation and-Air-Conditioning System, has 'oeen issued to provide instructions for
, periodic inspection of Hardcast tape. Maintenance Procedure H.168 is I
being prepared to provide detailed guidance in the repair of Hardcast tape.
The warranty period for Dow Corning 732 sealant is 10 yehrs. However, a SHUD procedure for periodic inspection, repair, and/or replacement of the sealant has not been written.
In summary, these sealing materials do not meet regulatory requirements, are not yet qualified, and are not covered by all of the inspection and maintenance procedures required for degradable materials.
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REQUEST FOR INFORMATION (RI)
.RI NO: 040 SYSTEM CODE: CR/TSC HVS ISSUE DATE: 07-28-87
SUBJECT:
- CR/TSC ESSENTIAL HVAC COMPONENT REMOVAL
-DEPARTMENT: MAINTENANCE COORDINATOR: JIM DARKE TEAM MEMBER: BRIAN A. KN0X TEAM LEADER; EFPEN ESPERANZA POTENTIAL CONCERN /0VESTION:
. Few provisions have been made for the removal and replacement of heavy HVAC components on the Auxiliary Building roof.
~ Section 3.13.3 of the Rancho Seco Technical Specifications requires the plant to go to hot standby mode after the following:
- One CR/TSC essential HVAC train is inoperable for more than 7 days.
- Both CR/TSC essential HVAC trains are inoperable for more than 3.5 days.
At present there are no permanent cranes, hoists or monorails on the Auxiliary Building roof to facilitate the removal of essential HVAC !
1 compressors,- motors, fans or coils. This problem was identified in the CR/TSC Essential HVAC System Status Report (Problem 9) which states "The refrigeration system arrangement does not accommodate convenient or efficient maintenance of the system." The absence of rigging equipment for the essential HVAC components will increase the duration of' major repairs and could cause a reduction of plant; generating time.
A draft maintenance procedure:(M.160) has been written for removal and reinstallation of the reciprocating compressors and their motors
.(U-545A-C and U-545B-C). However, this draft procedure does not address the following needs:
- a. There are no monorails, jib cranes or platforms to move heavy components from their housings to the edge of the roof,
- b. The procedure doesn't provide instruction for removing essential fans, fan motors and coils.
- c. The procedure states that the compressors and compressor motors can be replaced with the on-site Bantam crane. With boom fully extended
'it will not be able to lift the condenser coils.
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RI-040 (Continued)
The maxth:um reach of the Bantam crane boom is 75 ft. at a 16 degree 1 angle.. The block and hook makes it approximately 73 ft. The roof is 60 J ft. high with a 4 ft. parapet which makes the height of the wall ]
approximately 64 ft. This leaves only 9 ft. for rigging (slings, choker, j etc ) and load. At that height, the load limit is reduced to 5300 lbs 1 off the front of the Bantam crane and 3100 lbs off the back of the Bantam 1 crane. 1 i
-The only. accessible place to make a " pick" is the east side of the J stairwell between the Auxiliary Building and the T&R Building. This would necessitate making the " pick" from the front of the Bantam' crane.
Swinging the load to the right over the Auxiliary Building and setting it behind the Bantam crane makes the maximum load only 3100 lbs (1-1/2 tons). Renting and setting up a larger crane would increase the repair.
time by days.
In summary, the absence of adequate rigging equipment on the Auxiliary Building roof and procedures for its use, could cause a plant shutdown.
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REQUEST FOR INFORMATION (RI)
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RI NO: 041 SYSTEM CODE: CR/TSC HVS ISSUE DATE: 07-27-87
SUBJECT:
SETPOINT FOR PDISH-54703B/PDISH-547048 DEPARTMENT:- SYSTEM ENGINEERS COORDINATOR: JOHN ITTNER TEAM HEMBER: M. RICHARD TEAM LEADER: EFREN ESPERANZA POTENTIAL CONCERN /00ESTION:
Technical' Specification 4.10.1.B.4 requires that '5..the pressure drop across the Combined HEPA filters and charcoal absorber banks is
< 6 inches water gauge. . . .". The pressure switches for total filter' differential pressure for SF-A-7A/B (PDISH-54703B and PDISH-54704B) are set at 8 inches water gauge, e 2% (Process Standards for HVAC). These switches will not. alarm in the Control Room (computer points P2904/P2905) until the differential exceeds the Technical Specification limit by 2-inches water gauge.
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-(, -REQUEST FOR INFORMATION (RI)
RI NO: 042 SYSTEM CODE: CR/TSC HVS ISSUE DATE: 07-27-87
SUBJECT:
CONTROL ROOM /TSC ESSENTIAL HVAC OPERATING STATUS INDICATION IN THE CONTROL ROOM IS NOT ADEOUATE DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR: R. LAHRENCE TEAM MEMBER: M. RICHARD TEAM LEADER: EFREN ESPERANZA POTENTIAL CONCERN /0UESTION:
During normal plant operations, the essential'CR/TSC HVAC is required to be operable (Technical Specifications 3.13 and 4.10) or the plant must be shutdown within specified time limits. During review of this system, two conditions were identified that could affect operability,.but which were not annunciated or indicated in the Control Room. One condition was a power interruption to the control circuit for the condensing unit and the other was a failure of the compressor crankcase heater. System Status Report Problem 39 identified the CR/TSC Essential HVAC controls and operating status indication in the Control Room as being inadequate.
INTERRUPTION OF CONTROL CIRCUIT POWER i
Any interruption of power to the control circuit for the CR/TSC Essential HVAC condensing unit for any reason will render the conderising unit "not ready for auto start" until a manual reset button at the compressor is reset by an operator. When an interruption of power occurs, the 3CR relay (Ref: E-206 Sh. 154/155, DCN Sh. 3 of 3 for ECN R-0904D) requires manual reset to energize this relay prict to any auto start of the unit on Hi temperature, Hi rad or toxic gas mode. There is no local or control room annunciation / indication to inform the operator that the 3CR relay is not energized and, therefore, that the unit is "not ready for autostart" (to indicate to the operator the need to reset this relay).
This lack of annunciation / indication could result in the unit not being ready to perform its intended function.
FAILURE OF CRANKCASE HEATER Operating Procedure A.14 (limits and precautions) states that "the 1 crankcase heaters for U-545A/B must be '0N' for a minimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to compressor startup if the crankcase heaters were de-energized for one hour or longer." A compressor crankcase heater circuit is energized any time the control power is 'ON' and the compressor is not running (Ref: E-204, Sh. 154/155).
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. RI-042 (Continued)-
4
. Since-the condensing unit is normally in a standby status (ready.to start on Hi temperature, Hi Rad or toxic gas signal), the heater circuit would normally be energized. Since there is no annunciation for heater failure (or blown heater fuse), such a failure could go undetected while the condensing unit is in a standby status. With a failed heater, the unit '
could still start upon receipt of the start signal and operate with low oil pressure and excessive bearing wear due to lubrication with oil which contains refrigerant. The heater appears to be a " Required Auxiliary" (Technical Specification 1.3) for the Essential CR/TSC HVAC system, so a means'to detect failure should be available.
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k REQUEST FOR INFORMATION (RI)
RI NO: 045 SYSTEM CODE: CR/TSC HVS ISSUE CATE: 07-27-87
SUBJECT:
MCR/TSC ESSENTIAL FILTRATION UNIT DESIGN DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR: RON LAHRENCE TEAM MEMBER: J. S. GRILL TEAM LEADER: EFREN ESPERANZA POTENTIAL CONCERN /0VESTION:
The moisture separator sections in units SF-A-7A and SF-A-7B are not instrumented in accordance with ANSI N509-1980, or drained in accordance with ERDA 76.21.
SMUD Design Criteria 5108.10, Section 5.4.1 states that the safety-related air filtration units shall be designed in accordance with the requirements;of Reg. Guide 1.52, Rev. 2, ERDA 76-21, and ANSI N509.
Essential filtration units SF-A-7A and SF-A-7B each have a combination moisture separator /prefilter bank in the housing. A magnehelic differential pressure gauge connected across the separator /prefilter bank is provided. Two 2-inch external drain connections are supplied for the separator /prefilter section, but they are sealed with blind flanges. g Reg. Guide 1.52, Section C.2.g, states that "the ESF atmosphere cleanup system should be instrumented to signal, alard, and record pertinent pressure drops and flow rates at the control room." Table 4-1 of ANSI N509-1980 requires the following instrumentation:
- Local
- Demister/prefilter pressure differential indication
- Prefilter high pressure differential alarm l
. Remote Manned Control Panel
- Summation pressure differential alarm which includes the demister/prefilter differential Surveillance Test Procedures SP.84A and SP.848 require that the pressure l differential across the moisture eliminators be logged monthly. However, the acceptable pressure drop (0.5 inches w.g.) is not provided in the procedure. Bechtel Specification M13.16 lists the maximum permissible pressure drop as 0.5 inches w.g.
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RI-045 .(Continued)
Reg. Guide 1.52, Section C.3.h, states that " water drains should be
- designed in accordance with the recommendations of ERDA 76-21." ANSI N509-1980,- Section 5.4, states that the moisture separators shall remove 991, by weight of the moisture in an airstream containing 1.5 to.2 pounds of entrained uter per 1,000 cu. ft. of air. Bid Request M13.16, Section 9.6.2, states that the moisture separator shall remove water particles larger than 1 micron and shall collect and pipe the condensate to the.
housing drain. 1 The ASHRAE Handbook of Fundamentals indicates the particle diameter of fog ranges from 2 to 70 microns. The weight of the free moisture in fog common to the site was not provided in Specification M13.16. Some of the free moisture in the makeup airstream would be evaporated by the return air before reaching the moisture separator.
Bechtel P&ID M-504, Rev.13, indicates the moisture separator drains as capped with blind flanges. However, vendor print M13.16-1, Rev. O, indicates the drains uncapped. No notes are provided on P&ID M-504, in the System Design Bases (DB-HVS-5433A), or in the Design Basis Report for ECN A-3920A, B, C to explain why these drains should be capped.
In summary, present instrumentation and surveillance may not detect a dirty moisture separator /prefilter. Also, existing design bas'is y documents do not justify capping the moisture separator sump drains.
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REQUEST FOR INFORMATION (RI)
RI NO: 059 SYSTEM CODE: CR/TSC HVS ISSUE DATE: 07-28-87
SUBJECT:
TRAINING OF BUILDING MAINTENANCE HVAC PERSONNEL IN ELECTRICAL HORK DEPARTHENT: MAINTENANCE COORDINATOR: JIM DARKE TEAM HEMBER: JOEL HESTVOLD/ TEAM LEADER: EFREN ESPERANZA BRIAN KN0X POTENTIAL CONCERN /00ESTION:
A review of training records' indicates that Building Maintenance (BM)
HVAC personnel have not received any~ formal training in electrical circuitry even thoedi they are currently allowed to work on electrical control circuitry below 480 volts. This apparent lack of training could affect'the quality of QA Class I CR/TSC HVAC maintenance that has been performed on electrical circuitry.
A review of Hork Request (HR) #127110 and a draft of procedure M.160 seems to indicate that BH HVAC personnel are working on 480 volt power circuitry. This appears to be in violation of an informal agreement between BH and Electrical Maintenance (EM) which requires EM to do all .
power circuitry work of or above 480 volts.
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REQUEST FOR INFORMATION (RI) !
RI N0: 060 SYSTEM CODE: CR/TSC HVS ISSUE DATE: 07-28-87 I
SUBJECT:
CLQSURE AND RELEASE OF ECN R-1260 .
1 DEPARTMENT: SYSTEM ENGINEERING COORDINATOR: JOHN ITTNER TEAM MEMBER: J0EL HESTVOLD/ TEAM LEADER: EFREN ESPERANZA HOE RICHARD POTENTIAL CONCERN /0VESTION:
The following apparent discrepancies were identified in the turnover, release and closure of ECNs R-0938 and NCR S-4761.
- 1) The ECH turnover / release packages are not being stored as QA records per the~ requirements of AP.44 and RSAP-0601. The packages are not being stored in fire proof cabinets prior to turnover to the Records Information Center (RIC).
required actions. The turnover package and ECN R-0958 identify the need to address pressure tap installation in SP.618A/B. The ECN has been closed but SP.618A/B has not yet been issued. The turnover l package and ECN punchlist for ECN R-1260 identify the need to address trap modifications in SP.84A/B and RT-HVS-Oll. The ECN has been closed but the procedures have not lieen revised. NCR S-4761 required nixing uniformity testing per STP.1063A/B. The NCR has been closed but the STP.1063B test results were rejected by the Test Review Group and retesting is necessary.
- 3) A walkdown of the pressure tap installation for ECN R-0938 identified that the taps are not labeled. This may cause operator confusion during testing.
- 4) The revision to E-638, Sh. 7A to incorporate DCN 1A for ECH R-0938 mistakenly addresses the calculation for Section D, E, and F and Detail 5 as 2-FPR-I-0482 instead of Z-FPP-E0482. The minimum thickness of Dow Corning 3-6548 Silicone Foam for Section F was mistakenly identified as 9" thick instead of 11" thick.
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REQUEST FOR INFORMATION (RI)
RI NO: 065 SYSTEM CODE: CR/TSC HVS ISSUE DATE: 07-28-87 ,
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SUBJECT:
PATH FOR POTENTIAL INFILTRATION INTO TSC j DEPARTMENT: NUCLEAR-ENGINEERING COORDINATOR: RON LAHRENCE TEAM HEMBER: B0B THOMAS TEAM LEADER: EFREN ESPERANZA POTENTIAL CONCERN /00ESTION:
The Technical Support Center (TSC) is not maintained at a positive pressure relative to one of the surrounding air spaces during the high radiation operation mode.
The System Design Bases (DB-HVS-5433A) states that in the " radiological mode" the essential HVAC system shall maintain the control room (CR) and the TSC at a positive pressure of at least 0.125 inch water gauge (IWG)
-relative to the highest pressure adjacent to the CR/TSC boundary.
In the high radiation operation mode, the essential isolation dampers in the makeup air ducts are opened and the CR/TSC spaces are pressurized by one essential filtration fan.
When the CR/TSC Essential HVAC System is actuated, the TSC Normal HVAC Air Handler (AH-A-2) continues to operate. Static pressure in the duct on the Auxiliary Building side of the TSC isolation dantpers HV-54717 and HV-54718 is greater than that on the TSC side (0.14 ING differential when in the radiological mode 0.45 ING in the toxic gas mode). As a consequence, any leakage past these isolation dampers will result in unfiltered infiltration into the TSC.
The CR/TSC habitability study completed in July 1987 assumed 100 cfm (pressurized) to 110 cfm (isolated) infiltration leakage rates into the control room. However, isolation dampers HV-54717 and HV-54718 are not periodically leak tested to quantify their leak rate during emergency operation.
In summary, the TSC operating pressure in the radiological mode does not meet the system design basis and the rate of inleakage may cause the total inleakage to exceed the amounts assumed for dose and chlorine l concentration analyses. ,
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h .~, REQUEST FOR INFORMATION (RI)
RI NO: 070 SYSTEM CODE: CR/TSC HVS ISSUE DATE: 07-28-87
SUBJECT:
CR/TSC ESSENTIAL HVAC ISOLATION DAMPERS
-DEPARTMENT: MAINTENANCE COORDINATOR; JIM DARKE TEAM MEMBER: BOB THOMAS TEAM LEADER: EFREN ESPERANZA POTENTIAL CONCERN /00ESTION:
The CR/TSC Essential HVAC isolation dampers will not be able to close when they accumulate dirt. With low efficiency filters, dirt will accumulate, and no program could be identified -to insure they are maintained in a clean condition. This problem was identified in the CR/TSC System Status Report (Problem 19) which states " isolation dampers ;
not operating properly."
Tech Manual M13.14-472 for the CR/TSC Essential HVAC System isolation dampers includes recommendations to clean and lubricate the dampers on a periodic basis.
In an August 6, 1986, letter from J. Dodson to J. Naleway, the manufacturer further stated:
"There are no current standards for the performance of PAPC0 Bubble Tight Dampers when contaminated by dirt dr dust. Damper can operate i in a dirty humid environment. They have performed the isolation function even when badly contaminated, but this has been in violation of the design conditions, and cannot be used as a basis for establishing any sort of performance level."
Proper operation of the CR/TSC Essential HVAC dampers is essential to provide a positive 0.125 inches water gauge relative to the surrounding I atmosphere (Technical Specification 4.10.1.B.7).
In summary, the reliability of the isolation dampers for the CR/TSC is f questionable because of heavy dust loading and an inadequate maintenance program.
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i REQUEST FOR INFORMATION (RI) 4 RI.NO: 075 SYSTEM CODE: CR/TSC HVS ISSUE DATE: 07-29-87
SUBJECT:
ENVIRONMENTAL QUALIFICATION OF HY-54727 DEPAkiMENT: MAINTENANCE COORDINATOR: JIM DARKE TEAM MEMBER: J0EL HESTVOLD TEAM LEADER: EFREN ESPERANZA POTENTIAL CONCERN /00ESTION: ,
The lack of replacement of a solenoid at the time scheduled and the difference in replacement periodicity between MIMS and MARSS appear to be '
violations of the EQ program. !
ERPT E-0177 identifies HY-54727 as necessary to be qualified per !
10CFR50.49. To maintain this qualification, the Maintenance and Replacement Schedule Summary (MARSS) requires gasket replacement every 1.32 years with'the first replacement due 8/86. The Nuclear Engineering Group has stated that rather than simply replace the gasket, the whole !
solenoid would be replaced. The MIMS system identifies a PM task to replace the solenoid every 1460 days with the first replacement scheduled for 12-01-86. MIMS does not identify that this replacement has yet been '
accomplished. )
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6.0 ATTACHMENTS 6.1 List oT Documents Reviewed j 6.2 Status of RIs I i
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6.1 LIST OF DOCUMENTS REVIEWED System Design' Bases, DB-HVS-5433A, HVAC Systems for the Control Room and
' Technical Support Center, 07-03-87
- SHUD Updated Safety Analysis Report, through Amendment 4, 10-01-86, Sections 1.5.15 and 9.7
. Nuclear Engineering Design Criteria Manual, 06-29-87, Sections 5101.1 and
'5108.10.
Bid Request M13.16, Radiological Filtration Units, Class I, 01-27-82 Bechtel P&I Diagram M-504, Rev. 13, 05-01-87 SMUD P&I Diagram, M-504, Sheets 1, 2 and 3, Preliminary Control Room / Technical Support Center Essential Air Conditioning System, System Status Report, Rev. 1, 12-05-86 Control Room Habitability Study, Rev. 5, 07-87 Design Basis Report based on ECN R-0904, Rev. O, 04-18-87 Design Basis Report based on ECN R-1402, Rev. O, 06-30-87; CR/TSC ,
Essential HVAC Noise Reduction Design Basis Report based on ECN R-0769, Rev.\l, 06-22-87 Process Standards Manual PSM 185, 07-03-87; Section AP.154. Rev. 12, 10-07-86 Surveillance Procedure Manual, Rev. 26, 07-09-87, Procedures SP.84A/B, SP.485A/B, SP.618A/B (draft), SP.619A/B (draft)
Design Basis Report based on ECN A-3660Z, Rev.1,10-86 Design Basis Report for ECN A-3920A,B&C, 11-01-82 Elementary Diagram E-206, Sh. 154, Rev. 7c for ECN R-0904D, 05-12-87, Control Room and TSC HVAC Administrative Procedure AP.23.15, Rev. O, 03-07-87 Equipment Labeling Operating Procedure A.14, Rev. 25, 07-14-86, HVAC System Operating Procedure A.14C, Draft, 05-24-87, CR/TSC Building HVAC System Casualty Procedure C.51, Draft, 06-22-87, Loss of CR/TSC Ventilation ATTACHMENT 1
LIST OF DOCUMENTS REVIEWED (Continued)
System Training Manual, Rev. O, 10-18-84, Chapter 11 Lesson Plan OD 24 D 3300, Rev. O, 12-24-86 Operator Tour Inspection Sheets, HP1636B/D-00818, undated Rancho Seco Technical Specifications, Rev. TSH-76, 05-12-87, Sections 3.13 and 4.10, and Proposed Amendment 161 to Section 4.10 SMUD Maintenance Manual M13.16-IM01 Vendor Print M13.16-1, Rev. O, AAL 574, House Assembly, Filter Vendor Print AAF 1327352, Sheet 1 of 5, Filtration Unit Instrumentation i
Bechtel Letter BSL-4403, 12-19-84, Hardcast Tape HVAC Maintenance Procedure, M.111,' Air Balance of Ventilation Systems Preventive Maintenance Procedure, H.148, Cleaning Nucon BT and BTR Isolation Dampers !
Special Test P ocedures STP.1059 and STP.1061, Operational Verification ,.
of the Refrigeration System for the CR/TSC Essential Air System Trains.
A/8, and STP.1063A&B
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Routine Test Procedures RT-HVS-005, RT-HVS-005 and RT-HVS-011 Nonconformance Reports (NCR) S-4761, Rev. 1; S-5611, S-5663, S-64B7,
'S-6576 and S-6634 Occurrence Description Report (ODR)86-279 and 86-280 Engineering Calculation MARSS file Z-EQP-E0068 and Z-EQP-E0074 Engineering Change Notices (ECN) R-0938 (including turnover / release packages DBR, SAR, DVR) and R-1260 (including packages D8R, SAR, DVR)
Hork Requests (hrs) 115770, 118907, 120454, 122339, 122340, 126236, 126237, 126814, 126815, 127110, 127111, 127112, 127114, 127130, 128628, 128452, 131284, 131285, 132621, 135252 Preventive' Maintenance Work Request 68888 ATTACHMENT 1 i
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- $ LIST OF DOCUMENTS REVIEHED (Continued)
Drawing Change Notices (DCNs)
M-421.,Sh. 5, R7, DCN SA; M-421, Sh.:6 R5, DCN 3A; M-504,.RIO, DCN'19; I-1157, Sh. 5, R0, DCN 5A; C-954, Sh. C, R0, DCN_2; E-641,LR1, DCN SA; E-638, Sh.-7. RO, DCN 1A;
^
E-638, Sh. 7A,~,.R0,'DCN 2A; E-639, Sh. 1, R0, DCN 2A; E-639, Sh. 2. R0, DCN.1A; C-955, R0, DCN 3A;-
C-955, SH. K,.R0, DCN 3; and M-504,.RIO, DCN 17
-QA Audits 87-10, 87-21, 87-23 and 87-24 QA Surveillance Reports 809 and 87-5036 Engineering Report _E-0177 Field Problem Reports 1 and 2 to ECN R-0938 l
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ATTACHMENT 1
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j . } ;* 6.2 STATUS OF RIs r
RI NUMBER STATUS RSL NUMBER 006 OPEN 009 ACKNOWLEDGE RSL-RI009 013 OPEN 040 OPEN ,
041 OPEN 042 OPEN l 045 OPEN 059 OPEN 065 OPEN 070 OPEN 075 OPEN 4
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ATTACHMENT 2
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MT '. ' EXPANDED AUGMENTED SYSTEM REVIEH AND TEST PROGRAM (EXPANDED ASRTP)
I EVALUATION OF THE DECAY HEAT REH0 VAL SYSTEM SUBMITTED BY: ._/_> ~ ~
M I M 4 4 -- DATE: ' OB -e5 -87 EITH PRINCE
'[TEAMLEADER or vi1 w
.g
[ DATE: hd ' 6-) '
CONCURRENCE:/ DAVID HUMENANSKY EXPANDED ASRTP PROGRAM MANAGER
[
CONCURRENCE; / DATE:
/
f7 ' '
/ B0B'CROLEY \
DIRECTOR, NUCLEAR CHNICAL SF.RVICES
.j M L' ! ? O ! f G - - -
! '41 f'q 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 DETAILED OBSERVATIONS - REQUESTS FOR INFORMATION 9 6.0 ATTACHMENTS 28 6.1 List of Documents Reviewed 29 6.2 Status of RIs 31 ;
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9 M EXPANDED AUGMENTED SYSTEM REVIEH AND TEST PROGRAM w . EVALUATION OF THE DECAY HEAT REMOVAL SYSTEM
1.0 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 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 implemented on 33 selected systems and an in-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 Each multi-disciplined team consists of dedicated personnel ASRTP.
with appropriate backgrounds to evaluate the operations, maintenance, engineering, and design functional areas.
Independence, perspective, and industry standards provided by team members with consultant, architect engineer and vendor backgrounds are joined with the specific plant knowledge of SMUD team members.
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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 enhance the evaluation. Each team prepares a final report '
for each completed selected system evaluated. This report is for the Decay Heat Removal system.
4 s -:e; 2.0 : EUAP031-The objectives of the Expanded ASRTP evaluation areLto:(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 Decay Heat Removal 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.
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- 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 l as implemented for the system in the following functional areas:
- 1. Systems Design and Change Control
- 2. Maintenance \ i
- 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 1.
The primary source of leads for the team were the problems identified in the Decay Heat Removal 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 Decay Heat Removal system included a review of pertinent portions of support systems that must be functional in order for the Decay Heat Removal system to meet its design objectives.
s 4.0 'OVERALL RESULTS AND CONCLUSIONS
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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. Section 5.0 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. The numbers in brackets after each individual summary refer to the corresponding RIs in Section 5.0.
4.1 Summary of Significant Findinos 4.1. 1 Discussion Listed below is a summary of significant findings in the Expanded Augmented System Review and Test program (EASRTP) evaluation of_the Decay Heat Removal (DHR) System. Many 4 documents (see Attachment 1) were reviewed during this evaluation. The initiating document for the evaluation, however, was the System Status Report (SSR). Thi rty-five (35) problems were identified in the DHR System SSR. Each of the identified problems were reviewed and pursued until avenues of probe were exhausted or until additional concerns i' were identified. Only five (5) additional concerns were identified as a result of SSR pursuit which indicates that the SSR was generally effective.
4.2 Concern Assessment l
No one concern on its merit alone appears to challenge the intended functionality of the DHR System. There are a number of concerns, however, which when considered together, led the team to consider that the total DHR System may not be as dependable as it should be. These concerns include:
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- Possibility for system overpressurization i
= Water hammer problems l
= Reported excessive pipe radial movement l . Possible to violate successive start limits of DHP
- Possible vortexing problems when DHP takes suction from BHST
= Circuitry design affects system reliability
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Consideration did not lead to determination that DHR System j was less than dependable. It is apparent, however, that the DHR System has problems and potential problems that require immediate attention and follow through to correction.
(RI-23) (RI-52) (RI-54) (RI-67) (RI-76) ;
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'A- OVERALL RESULTS AND CONCLUSIONS (Continued)
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4.3 No Jacketina on Emercency Sumo Recirc Pioina (Source:
Halkdown)
System walkdown identified the absence of-protective jackets around the Emergency Sump Recirc Piping in the DH Pump rooms. Also, the Emergency Sump Isolation valves are jacketed (partially), but are not leak tight. The as-built design does not conform to B&W Design Basis Document, original QA Manual Code application,'nor ANSI N-271 Requirements For Containment Isolation. (RI-18) 4.4 Potential for Water Hammer (Source: SSR Problem 20)
High point _ vents are not installed in "A" DHR System. This prevents system from being properly vented and allows potential for " water hammer" and associated damage upon each system initiation. Continuing problems with water hammer in a system is indicative of a design deficiency and a significant safety issue. (RI-23) 4.5 Potential for Floodina DH Pumo Rooms (Source: USAR)
DH Pump rooms are not isolated by a water tight door as stated in USAR. In case of-line break, DH Pump room will g contain approximately 60,000 gallons of water before spilling into "B" DH Pump room. Action to mitigate consequences of a pipe break must be taken after receipt of level alarm indication. Level switches and sump pumps are neither Safety-Related nor EQ Qualified. Level indicators are powered by Non-Q "E" Bus. (RI-44) 4.6 Testina and Calculations 4.6.1 The team identified several tests or calculations required, but not in place, to determine system capability at specified conditions or to ensure system operability.
. SP.203.05A and B, as written, do not ensure that suction for the test comes only from the SF Pool.
(Single Source) (RI-02)
- DHR Syster SSR does not provide for testing DHR Pump to Pressurizer Auxiliary Spray. The function is provided in DHR System SSR, Rev.1, Section 2.2.1.4. (RI-03)
. Testing DH Pump to HPI (piggyback mode) does not cover HPI and Hakeup Pump running in parallel. Need to determine adequate NPSH available. (RI-72)
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OVERALL RESULTS AND CONCLUSIONS (Continued) l:
4.6 (Continued)
. Calculations are required (currently being performed) 3 to determine if vortexing could occur when DHR Pump takes. suction from BHST at low levels. Calculations are available for same potential problem with Emergency Sump. (RI-76) 4.7 Generic Concern 4.7.1 Velan valves in the DHR System were found to contain carbon steel bolts and are subject to corrosion and failure.
Moreover, many valves in the following systems are known to have carbon steel bolts and are subject to the same boron.
corrosion failure:
RCS PLS BHS RCD i SIM- CBS SfC RHS The carbon steel / boric acid corrosion problem was recognized
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at Rancho Seco:Es early as 1979. Since that time ECNs A-2921 and A-2931 were issued to replace cart,on steel bolts T, .with.630 S/Suin a number of Anchar Darling and Velan valves. DHR System Anchor valves had bolting replaced, Velan valves in DHR System had no replacement. A brief walkdown of the affected systems will indicate the magnitude of the problem. (RI-68) \
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', i a 5.0 DETAILED OBSERVATIONS - RE00EST FOR INFORMATION 9 , .
- 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.
Tbsse RIs are considered drafts throughout the. entire evaluation anti.l.they become part of the final report. Responsible '
organizations can accept the potential concern as valid or they may di agree with the potential concern. If they disagree, they can
, submit information that convinces the EASRTP team members that the potential concern is nct valid, or they may redirect the-EASRTP mer.bers to better focus the concern. RIs developed during the sjstem evaluation comprise'this:section of the report.
Attachment 2 of the report 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.
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Approximately one week will be provided.after the report is issued to provide time for departments to address each RI for validity. A g revision to Attachment 2 will then be issued to reflect the status of RIs. All RIs not acknowledged at the end of this period will have an "Oper," status. RIs are then transferred into the Restart Scope List track',ng system for resolution and corrective action implementation.
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4 2 REQUEST FOR INFORMATION (RI)
Y 07-22-87
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'RI NO: 002 SYSTEM CODE: DHS ISSUE DATE:
SUBJECT:
- TESTING OF BACKUP SPENT FUEL COOLING MODE USING DECAY HEAT PUMP DEPARTMENT: OPERATIONS ' COORDINATOR: RICH MACIAS TEAM MEMBER: DS. DENNIS MARTIN.- TEAM LEADER: KEITH PRINCE S. CARMICHAEL POTENTIAL CONCERN /00ESTION:
T There is a possibility of meeting the test acceptance criteria without meeting design basis flow requirement of 3000PM from a single source (BHST/or spent fuel pool).
SP.203.0EA and B, Revisien 19, does' not require. recording the spent fuel pool levet prior to and following the recirculation test for verification of suction flow being 3000 GPM from spent fuel pool. There is a possibility of taking suction from another source which does not meet the intent of the surveillance procedure. There is no suction flow indication in the DHS for either the BHST.or the spent fuel pool. - The flow is indicated only at discharge of DH cool rs by FI-26003, FI-26004, FI-26048A, and FI-26049A.
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F REQUEST FOR INFORMATION (RI)
RI NO: 003 SYSTEM CODE: CHS ISSUE DATE: 07-18-87
SUBJECT:
DECAY HEAT SYSTEM AUXILIARY SPRAY TESTING DEPARTMENT: NUC OPERATIONS COORDINATOR:. JOHN ITTNER (SYSTEMS ENGINEER) m
. TEAM MEMBER: D. SAT 0ATHY TEAM LEADER: KEITH PRINCE S. CARMICHAEL POTENTIAL CONCERN /00ESTION:
The function of auxiliary spray is provided in the DH Removal SSR, Revision 1, Section 2.2.1.4. However, this alternate method of pressurizer cooldown is not mentioned in any testing section of the SSR.
How this function is verified is not addressed e;g., by taking credit from tests or by other means. .This function is not presently addressed in the DHR SSR' testing section.
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, +9 REQUEST FOR INFORMATION (RI)
RI NO: 012 SYSTEM CODE: DHS ISSUE DATE: 07-20-87
SUBJECT:
DISCREPANCY IN LIMITS AND PRECAUTIONS'BETHEEN VARIOUS PROCEDURES DEPARTMENT: OPERATIONS COORDINATOR: RICH MACIAS TEAM MEMBER: D. MARTIN TEAM LEADER: . KEITH PRINCE
. POTENTIAL CONCERN /0UESTION:
Review of Nuclear Operations Procedures for the Decay Heat System indicates that the limits for the same operation conflict from one procedure to the next. Some examples are:
PROCEDURE NO. / PARAMETER
- Excessive flow rates on DHS pump while lined up to take suction from the emergency sump increases the potential for loss of NPSH g and Vortex Formation
- 2. Maximum successive starts \
for DH Pump Motor 0 Ambient Temp: A.8/2 AP.23.10/3
- 3. Maximum successive starts for DH Pump motor @ Hot: A.8/ AP.23.10/2 Several successive starts of a large motor can cause stator damage from heat buildup caused by high starting current without sufficient time for the motor to cool.
- When indicated RCS Pressure is 290 psig the pressure at the suction of the DH Pump is greater 3
than the 300 psig design pressure of the suction piping.
3 REQUEST FOR INFORMATION (RI)
- RI NO: 016 SYSTEM CODE: RCS ISSUE DATE: 07-22-87
SUBJECT:
REACTOR COOLANT PUMP COMBINATIONS DURING C00LDOWN DEPARTMENT: NUCLEAR OPERATIONS COORDINATOR: RICH MACIAS TEAM MEMBER: D. SATPATHY TEAM LEADER: KEITH PRINCE S. CARMICHAEL POTENTIAL CONCERN /00ESTION:
Plant Cooldown Procedure B.4 calls for operation of one Reactor Coolant Pump per loop. The current B&W recommendation is to cooldown with two RCPs in one loop. This is because with one pump per loop the pumps are operating near .the runout condition where there is risk of cavitation damage, increased vibrations, and inadequate NPSH at low RC pressures.
The operating procedures at Davis-Besse and Crystal River have been revised to go into decay heat cooldown with 2/0 RC pump combination.
~ Moreover, there are added limits and precautions to. minimize. running a single RC pump per loop, i.e.,1/1 or 1/0 pump combination (to less than 5 minutes).
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REQUEST FOR INFORMATION (RI)
, . ]. 07-30-87 RI NO: 017 SYSTEM CODE: DHS ISSUE DATE:
SUBJECT:
DECAY HEAT REMOVAL COOLER CAPABILITIES DEPARTMENT: SYSTEM ENGINEERING COORDINATOR: JOHN ITTNER TEAM MEMBER: D. SATPATHY S. CARMICHAEL TEAM LEADER: KEITH PRINCE POTENTIAL CONCERN /0UESTION:
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Design bases document, Section 4.1 (2) under licensing design. bases requires each D.H. cooler be designed to cool the ECCS sump water. The containment building pressure analysis is based on this cooler performance. The SSR (0.H. Removal-System) rev. 1, page 4-15, paragraph-7, requires calculation of heat transfer co-efficient. It is more appropriate to use the actual cooler outlet water temperature as direct verification of the cooler performance and document the validity of the Decay Heat Removal Cooler Characteristics. This could replace the analytical method which may not be accurate due to.possible crud build up, blockage or degradation of the cooler performance.
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a REQUEST FOR INFORMATION (RI)
RI NO: 018 SYSTEM CODE: DHS ISSUE DATE: 07-22-87 i
SUBJECT:
JACKETING AROUND REACTOR BUILDING SUMP ISOLATION VALVES DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR: RON LAWRENCE i
TEAM HEMBER: S. CARMICHAEL TEAM LEADER: KEITH PRINCE POTENTIAL CONCERN /00ESTION:
Decay heat system design bases document Section 4.l(8) page 13 of 49 under Licensing Design Bases states, "A jacket shall be installed at the exterior of the Reactor Building. Jaclosed piping up to and including the stop valve in each DHS suction line from the Reactor building."
The USAR, Section 5.2.4, page 5.2-45 under Reactor Building Isolation states:
"Each of the two emergency sump recirculation lines has only one isolation valve, encased in a secondary housing, and located outside the Reactor Building. These valves are required to open under
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certain emergency conditions and can fail in two ways: (a) the valve body ruptures or leaks and (b) the valve fails to open. The [
housing takes care of the failure under (a) and 100 percent redundancy, provided by two recirculation lines, takes care of the failure under (b)." \
This requirement was provided by B&W in the 18K1 manual " Duke type PHR Nuclear Steam System" issued 05/15/68.
The jacket is provided to prevent drainage of the reactor building emergency sump in the event of failure of the stop valve body or breakage of the suction line between the Reactor building penetration and the valves.
During walkdown of the DH suction line, it was noted that the jacket around the valve body (HV-26105) is not leak tight. There is no jacket around the piping from valve HV-26105 to the penetration.
Without a proper designed jacket around the piping and the valve, there is a possibility of draining the entire emergency sump inventory during moderate line crack to the auxiliary building decay heat system and loss of lESH to the redundant decay heat pump.
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I REQUEST FOR INFORMATION (RI)
RI NO: 023 SYSTEM CODE: DHS ISSUE DATE: 07-22-87
SUBJECT:
DECAY HEAT SYSTEM LICENSING DESIGN BASES I
. DEPARTMENT: NUCLEAR ENGINEERING C0ORDINATOR: R. LAHRENCE-TEAM MEMBER: D. SATPATHY TEAM LEADER: KEITH PRINCE' DENNIS MARTIN POTENTIAL CONCERN /00ESTION:
The Design Bases document, Section 4.1 under Licensing Design Bases (Paragraph 13) requires the DHS to be completely filled at. initiation so that venting is not required. This criteria is from " Duke type PHR Nuclear Steam System" 18K1 manual issue 05-15-68.
However, no means has been provided to keep the DH system. filled with water. DH pump "A" discharge piping does not have high point vents and adequate drain locations to vent the system properly. .
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REQUEST FOR INFORMATION (RI)
RI N0: 038 SYSTEM CODE: GENERIC ISSUE DATE: 07-27-87
SUBJECT:
CONTROL OF VENDOR TECHNICAL MANUALS DEPARTMENT: ADMINISTRATIVE COORDINATOR: HARRY ELLER TEAM HEMBER: D. A. LOGAN TEAM LEADER: KEITH PRINCE D. SATPATHY POTENTIAL CONCERN /0UESTION:
Revision 4 of AP.46 dated 3-30-87 " Control of Vendor Technical Manuals" is not being fully implemented; training to AP.46 requirements has not been provided to appropriate personnel. (e.g. NEDC)
. The responsibilities imposed upon Nuclear Engineering Document Control (NEDC) as defined in Section 6.0 of AP.46 are actually being implemented by personnel in the Technical Library.
. Enclosure 8.1 of AP.46 " Technical Manual Review Sheet" is not being utilized as required by Section 6.2. Although, it should be noted that the review sheet currently used, requires essentially the same information as Enclosure 8.1.
. Distribution of technical manuals for review and approval (Ref. 6.3) is not by NEDC, but by the Technical Litirary.
- NEDC personnel, when queried of the requirements of AP.46 indicated that they have no knowledge of its contents. Similarly, Nuclear Procurement was not aware of the requirements imposed upon them (Ref. 6.10). It was indicated that they had no knowledge of the issuance of this AP.
. How does the Technical Library determine distribution of " approved manuals" (Ref. 6.5.3)? (Plant has not responded.)
. Have all Engineering / Design personnel (as applicable), or other personnel who would receive vendor technical documents, been made aware of AP.46 requirements to ensure vendor technical manuals are being updated? (Plant has not responded.)
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. g. RI-038- (Continued)'
. Why doesn't AP.46 make reference t0 AP.42 " Maintenance Information Management System" (MIMS) procedure?- Contained in the MIMs program is the Master Equipment List (MEL). The MEL, in part, provides reference to Manufacturer's Instruction Book No.'s.(Ref. AP.42, Rev. 5, pg. 40). Who is responsible for ensuring that Manufacturer's (vendor's) instructions, either new or revised, are incorporated into the MEL? Shouldn't AP.46 address who has this responsibility etc.? (Plant has not responded.)
. AP.46 does not address cross referencing of vendor manuals where information is contained in one or more than one manual or where one manual covers multiple equipment.
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REQUEST FOR INFORMATION (RI)
RI N0: 044 SYSTEM CODE: DHS ISSUE DATE: 07-27
SUBJECT:
FLOODING IN DECAY HEAT PUMP ROOMS
. DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR: R. LAWRENCE TEAM MEMBER: D. MARTIN. S. CARMICHAEL TEAM LEADER: KEITH PRINCE DILLAP SATPATHY POTENTIAL CONCERN /00ESTION:
USAR Table 14.1-17 Item 22, indicates that- a barrier is provided between the DH pump rooms to stop room-to-room flooding.
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Halkdown of the DH pump rooms determined'that the DH pump room "A" has a dimension of approximately 16 ft. x 60 ft. There is' a fire door at -39' elevation. As such, approximately 8000 CFT of water can be contained in the 'A' pump room before water spills over to the 'B' pump room. There are apparently no water tight doors between decay heat pump rooms, rather, fire doors have been provided. It appears that a potential deficiency exists in water tight separation between decay heat pump rooms as addressed in the USAR.
Flooding is isolated by operator action after receipt of DH ~pumpJroom I level alarms. These levels switches LSH-66407 and LSH-66311 ar( located on the walls of DH pump rooms,1 ft. above the floor. Neither the. level switches nor the sump pumps are safety related or EQ qualified. The ,
level switches are powered from the 'E' bus which is non-Q.
There is a possibility of flooding both DH pump rooms before operator action .is taken to mitigate it, as credit cannot be taken for non-Q equipment and the level switches may be damaged due to harsh environment. There is no level indication other than~ annunciation alarm in the control room.
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REQUEST FOR INFORMATION (RI)
RI NO: 052 SYSTEM CODE: DHS ' ISSUE DATE: 07-27-87
SUBJECT:
DECAY HEAT PUMP TRIP / RESTART LOGIC DEPARTMENT: TRAINING COORDINATOR: PAUL TURNER TEAM MEMBER: TOM FAUBLE TEAM LEADER: KEITH PRINCE POTENTIAL CONCERN /00ESTION:-
An Operator could violate successive start! limits on the Decay Heat Pumps by attempting to start the pump with 62/TD0 closed, not realizing the breaker closed when the start button was pushed.
Training materials for DHS state that the decay heat pumps cannot be restarted for 3 minutes if either of the suction valves (HV-20001 or HV-20002) leaves its open seat. Per drawing E-203 . Sheet 3, it appears that-the pump can be restarted as soon as both suction valves are full open regardless of the 3 minute time delay because the 62A/ INST contact would open de-energizing the trip coil.
Also, pushing the start button will start the pump whether the trip relay is initially energized or not. If this is done after 62A energizes and.
before 62A/TD0 opens, the pump breaker will retrip when the start button is released.
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REQUEST FOR INFORMATION (RI) l l RI NO: 054 SYSTEM CODE: DHS ISSUE DATE: 07-31-87 L
SUBJECT:
DHS SUCTION VALVE CIRCUITRY AFFECTS SYSTEM RELIABILITY-DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR: RON LAHRENCE TEAM MEMBER: THOMAS FAUBLE TEAM LEADER: KEITH PRINCE __
POTENTIAL CONCERN /0UESUJ)_fi:
The auto-close feature 'of Decay Heat Suction Valves HV-20001 and HV-20002 is a source of-system unreliability. Closure of either valve renders both DHS trains inoperable. Several reportable occurrences i
resulted from inadvertent closure of the SE valves (LERs 82-15, 86-16, I 86-24, 86-30,-and 87-38). The causes of the events were loss of SFAS A or B power; loss and subsequent restoration of power to HV-20001; and a lifted lead on the core flood tank valve HV-26514 interlock during
- maintenance ~on the CFT valve. Refer to elementary diagrams E203 sheets l
600, 60E for the following concerns: .
1 I . The way the circuitry is designed, the valves will auto-close on restoration of control power. The corrective action section of LER 86-30 says: .;
"A review will be performed concerning the necessity for
' auto-close interlocks (re-energizing the valve motor's MCC bus and having the valve remain in its last position) on HV-20001 and HV-20002 in light of the AE00 report on decay removal problems by July 6, 1987 (reference: AE00/C503
" Decay Heat Removal Problems at U.S. Pressurized Hater Reactor's dated December 1985). Any appropriate design changes deemed necessary by this review will be installed prior to the end of the next refueling outage."
What improvements are planned to solve this problem? (Plant has responded with plans to evaluate if design change required.)
. The amber light on the breaker cubicle is wired such that it will be lit if the pressure interlocks are met, but is not affected by the status of the CFT valve interlock. This is not consistent with a good human factors approach. The light, if it is to be useful, should indicate that all applicable interlocks have been met.
. The 62A/TDC contact that was installed by ECN A-2487 to override the pressure interlocks after a loss of SFAS A or B power also overrides the CFT valve interlock.
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-" RI-054'Rev. 1 (Continued)
' . Other B&W plants do not have the CFT valve interlock. Is this J
.. interlock required?'.(1)
. .What is,the basis for having an open-enable pushbutton at :the breaker cubicle? (Plant has not responded.)
- Can-the auto-close, feature be eliminated entirely by providing an alternative means of overpressure protection? (2)
-1 (Plant responded that interlock is required to prevent
'overpressurization of DHR system piping.)
2 (Plant responded that installation of other means of protection such
.as larger relief valves allows other: B&W plants not to have this interlock.)
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REQUEST FOR'INFORMATION (RI)
RI NO: 067 SYSTEM CODE: DHS ISSUE DATE: 07-29-87
SUBJECT:
LOW RANGE AND HIDE RANGE PRESSURE INSTRUMENT LOOPS DEPARTHENT: NUCLEAR ENGINEERING COORDINATOR: R. LAHRENCE and_ NUCLEAR OPERATIONS R. MACIAS TEAM HEMBERS: S. CARMICHAEL. D. SATPATHY. J. AREVALO. D. MARTIN ,
TEAM LEADER: KEITH PRINCE POTENTIAL CONCERN /0VESTION: .
-A potential system pressure error could cause DHR system relief valves to lift allowing RCS coolant to spill in Auxiliary Building.
The low range pressure indicator, PI-21261, provides RC pressure indication in the control room for decay heat system initiation. The 3 pressure indicator loop has an accuracy of e 10 psi. The wide range l pressure transmitters PT-21092 and 21099 are used for interlock permissiveness of DH dropline valves HV-20001 and 20002. These pressure transmitter loops have an accuracy of about z 25 psi. As such, there can be a difference of indicated RC pressure vs valve interlock pressure up to as much as 35 psi. This means the decay heat system may not be operable up to about 215 psi (250-35) which is less than RC pump NPSH' requirement for 1/0 RC pump operation, (Prodess Standard Curve AP.101-2B). DH System may be put in service when the RCS pressure is at about 285 psig (with potential 35 psig error) which violates the design pressure criteria of DH piping and equipment, i
c REQUEST FOR INFORMATION (RI)
.. RI N0: 068 SYSTEM CODE: DHS ISSUE DATE: 07-28-87
SUBJECT:
CARBON STEEL BOLTING MATERIAL r
DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR: R. LAWRENCE-TEAM MEMBER: TOM FAUBLE TEAM LEADER: KEITH PRINCE POTENTIAL CONCERN /00ESTION:
Carbon steel bolting materials are used in components that carry borated
-water. An IE Bulletin was recently issued concerning corrosion of carbon steel materials due to boric acid leakage. Small leakages over long periods of time can cause severe degradation causing fastener failure and breaching of the pressure boundary.
The problem was recognized about 1979 and ECNs A-2921 and A-2931 were issued to replace the carbon steel bolting on some Velan and Anchor '
valves, respectively, with Grade 630 stainless. However, not all affected valves received the replacement bolting. In the decay heat system, bolting was replaced on the Anchor valves in safety-related service, but not on Velan valves.
This concern is applicable to RCS, SIM, PLS, CBS, BHS, SFC, RCD and RHS.
All of these systems have valves with A-193 Grade B carbon steel bolting.
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-REQUEST FOR INFORMATION (RI)
RI-NO: 072 SYSTEM CODE: DHS ISSUE DATE: 07-29-87
SUBJECT:
TESTING OF HPI PIGGYBACK MODE J
. DEPARTMENT: SYSTEMS ENGINEERING C00R!,u ', (OR: R. LAWRENCE
'TENMMEMBER: TOM FAUBLE TEAM LEADER: KEITH PRINCE POTENTIAL CONCERN /00ESTION:
- Testing of the decay heat cooler to HPI suction line (piggyback mode) did not cover.-the case of an HPI pump and makeup pump operating in parallel.
, The original test, TP 203-4, demonstrated:the capability to deliver 528 gpm to the:HPI pump suction with sufficient NPSH available to the pump. However, with both an HPI pump and makeup pump operating, it is possible to get nearly twice as much flow and still be within the guidelines of the E0Ps (Rule 2).
It appears from the margin available in the original' test-that the
. capability exists, however, it has not been demonstrated by a test or o calculation.
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.a REQUEST FOR INFORMATION (RI)
RI NO: 076 SYSTEM CODE: DHS ISSUE DATE: 07-29-87 4
SUBJECT:
VORTEX FORMING AND NPSH FOR DH PUMPS V:
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DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR: R. LAWRENCE
- TEAM MEMBER: D. SATPATHY. TEAM LEADER: KEITH PRINCE S. CARMICHAEL POTENTIAL CONCERN /0UESTION:
The problem of vorte'ing x when DH pumps take suction from the Emergency Sump has been addressed. Similar analysis was not found, however, for DH
- Pumps taking suction from BHST at low levels. There is no assurance.that vortex formation will not occur when both DH Pumps operate simultaneously with Containment Spray Pumps prior to switch over to Emergency Sump suction. Also, there is no mention of importance of BHST level .
instrumentation accuracy and alarm indication used for operator action f for switch over. -
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k REQUEST FOR INFORMATION (RI)
RI NO: 077 SYSTEM CODE: DHS ISSUE DATE: 07-31-87
SUBJECT:
VIBRATION ON DECAY HEAT PIPING DEPARTHENT: NUCLEAR ENGINEERING COORDINATOR: R. LAWRENCE TEAM HEMBER: D. MARTIN TEAM LEADER: KEITH PRINCE S. CARMICHAEL l D. SATPATHY
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POTENTIAL CONCERN /00ESTION:
DH Cooler Bypass piping 26123-6"-GD is supported only by piping 26021-10"-GD and 26121-10"-GD. Plant operators have observed at as much as 3 inches radial movement of this piping.
The procedure for decay heat initiation is being revised to start the pumps with 'B' Bypass valve HV-26038, located in the middle of line 26123-6", open. There is potential for loads on this piping during DH initiation that may cause excessive movement without proper supports.
HV-26038 is a modulating valve controlled remotely from the control room without an operator assigned to the DH pump rooms. Problems that might be caused by such movement would not be immediately known. Discussions (
with the pipe support group indicated that proper supports will be provided in the long term as priority 3. Failure of this piping could affect ability to properly cooldown and maintain the plant in cold shutdown condition.
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6.0 ATTACHMENTS 6.1 List of Reviewed Documents 6.2 Status of RIs i
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- 9. ' o L e-LIST OF REVIEHED DOCUMENTS
- 1. System Design Bases for DHS, NEP 5419, Rev. 1 (draft)
- 2. System Status Report "DHS"..Rev. 1
- 3. USAR Section 1.4, 6.2, 9.5.2, 14.2.2
- 4. Technical Specifications 3.3, 3.8,=4.5.1, 4.5.3
- 5. Licensed Operator Training Program, OD 21 I 0800
- 6. Licensed Operator Training Program, 0D 21 I 0803
- 7. Plant Operating Procedures A.8, DHS, Rev. 30
- 8. Plant Operating Procedures B.2, Rev. 38' 9.. Plant Operating Procedures B.4, Rev. 40
- 10. Plant Operating Procedures B.6, Rev. 26 .
- 11. Plant Operating Procedures B.9, Rev. 12
- 12. Process Standards AP.103, Rev. 12
- 13. Casualty Procedure C.12, Rev. 5
- 14. Casualty Procedure CP.101, Rev. 4
- 15. Casualty Procedure AP.103, Rev. 4
- 16. Administrative Procedure AP.23
- 17. Administrative Procedure AP.42, Rev. 5
- 18. Administrative Procedure AP.44, Rev. 11
- 19. Administrative Procedure AP.48, Original !
- 20. Administrative Procedure AP.49, Original
- 21. Emergency Operating Procedures
-22. Maintenance Administrative Procedure MAP.0006
- 23. Code of Federal Regulations 10CFR50 Appendix.A.B,K,J 10CFR21-10CFR100 Appendix A
- 24. SYSTEMS Training Manual 10, Decay Heat' Cooling
- 25. SYSTEMS Training Manual 27, Emergency Core Cooling
- 26. SYSTEMS Training Manual 35, SFAS
- 27. Inservice Inspection Plan ISI 50-312
- 28. P&ID H.522, Sh 1
- 29. P&ID H.521, Sh 1,2,3
- 30. P&ID M.544
- 31. P&ID M.545
- 32. Isometric Drawings: All for DHS
- 33. ECN A-2921 A-2931 R-0498
- 34. Temporary Changes to Operating Procedure A.8, from 01-01-85 to l.
06-01-87 )
- 35. IDADS Computer Points for "DHS" Group Display 1
- 36. IDADS Computer Points for "SFAS" Group Display l
- 38. Licensee Event Report 86-30 87-38 ATTACHMENT 1
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LIST OF REVIEHED DOCUMENTS (Continued)
~39. Surveillance Procedures: 18 19, Rev. 0 29A 29B
, 203.05A, Rev. 19 203.058, Rev. 19.
203.06A,.Rev. 8 203.06B, Rev. 10 203.06C/D, Rev. 11 203.09, Rev. 10 203.11, Rev. 7 204.03A, Rev. 20 200.02 204.03B, Rev. 23.
213.01, Rev. 9
- ,- 214.01, Rev. 6 214.02, Rev. 2 214.03, Rev. 33
- 40. =NCR 4449 i l
NCR 6797 NCR S-006 i NCR S-029
- 41. Work Request.#134850. -
W/R 4776 H/R' 67 i H/R. 12552 \ l W/R 15609 l' H/R- 52036 H/R 80511 !
H/R 105763 !
H/R 98585 'l H/R 102132 q H/R' 102251 1 W/R. 102676
-W/R 100299 W/R 117875 4 H/R 129742 H/R 130460 H/R 133457 H/R 114607 i
ATTACHMENT 1
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STATUS OF RIs
- 1.
L RI NUMBER STATUS RSL NUMBER l -.
- 02. ACKN0HLEDGED RSL-RI-02 03 OPE!4 12 ACKNOWLEDGED RSL-RI-12 1
16 OPEN 17 , OPEN 18 ACKNOWLEDGED RSL-RI-18 23 ACKNOWLEDGED RSL-RI-23 29 ACKN0HLEDGED. RSL-RI-29 38 OPEN
- 44. ACKN0HLEDGED RSL-RI 52 OPEN
'54 ACKNOWLEDGED RSL-RI-54 67 ACKNOWLEDGED RSL-RI-67 68 ACKNOWLEDGED RSL-RI-68 .
72 OPEN.
76 ACKN0HLEDGED RSL-RI-76 77 ACKNOWLEDGED RSL-RI-77
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I ATTACHMENT 2 ;
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. v' EXPANDED AUGMENTED SYSTEM REVIEH AND TEST PROGRAM i (EXPANDED ASRTP)
EVALUATION OF THE EMERGENCY DIESEL GENERATOR SYSTEM i
SUBMITTED BY: e [ o N, NR // '
DATE: 8-/"8I TiiOMAS N. MARSfiALL TEAM LEADER CONCURRENCE: i % DATE: Tf.5-Y7
[DAVIDHUMENANSKY f
EXPANDED ASRTP PROGRAM MANAGER CONCURRENCE: , DATE:
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/ BBB CROLEY /\
DIRECTOR,NUCLEARTECHt/ICALSERVICES
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[ .. TABLE OF CONTENTS !
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Pace Number
'l.0 . INTRODUCTION 3 2.0'. PURPOSE 4
! 3.0 . SCOPE 5 4.0 OVERALL RESULTS AND CONCLUSIONS 6 5.0 DETAILED OBSERVATIONS REQUESTS FOR INFORMATION '9
- 6.0 ATTACHMENTS ~ 14 6.1 List of Documents Reviewed 15 6.2. Status of RIs 18 .
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4 EXPANDED AUGHENTED SYSTEM REVIEW AND TEST-PROGRAM j s
EVALUATION OF THE EMERGENCY DIESEL GENERATOR SYSTEM
1.0 INTRODUCTION
3 The Rancho Seco Expanded Augmented System Review and Test Program ]
[ASRTP) evaluation effort involves an assessment of the I effectiveness of the System Review and Test Program [SRTP] and an analysis of the adequacy of 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 l' implemented on 33 selected systems and an in-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 i members with consultants, architect engineer and vendor backgrounds are joined with the specific plant knowledge of SMUD team members.
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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 i 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 enhance the evaluation. Each team prepares a final report for each completed selected system evaluated. This report is for the Emergency Diesel Generator system.
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s 2.0 PURPOSE.
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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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f 3.0 SCOPE To accomplish the first objective, the Power Systems team evaluated the Emergency Diesel Generator 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) wat 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_ Power Systems team reviewed the programs as implemented for the Emergency Diesel Generator (EGS) .
system in the following functional areas: (-
- 1. Systerns Design and Change Control
- 2. Maintenance
- 3. Operations and Training
- 4. Surveillance and Inservice Testing
- 5. Quality Assurance
- 6. Engineering Programs i
The Power Systems team reviewed a number of documents in preparation for and during the Expanded ASRTP' evaluation of the EGS system.
This list of documents is found in Attachment 1.
The primary source of leads for the team were the problems identified in the Emergency Diesel Generators 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 EGS system included a review of pertinent portions of support systems that must be functional in order for the .
EGS 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. Section 5.0 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. The numbers in brackets after each individual summary refer to the corresponding RIs in Section 5.0.
4.1 Lack of a Fuse Control Program The inspection team has a concern that tha Emergency Diesel Generators may fail to perform their required function because of int.dequate fuse control.
4.1.1 A Fuse Control-Program should ensure that an adequate design review is performed, which addresses all fuse parameters and requirements for individual applications. A review of ECN's Ril50, R1060 and Ril2SE, all of which installed fuses, revealed a lack of adequate design review addressing the added fuses. On ECNs R1150 and R1060, no documentation could be found which addressed fuse coordination with other protective devices, interrupting rating, voltage rating, qualification for DC applications or required amperaga to
' support function. A calculation, Z-EDS-E0671, was fourd #or h ECN R1128. It addressed the required parameters in most cases. It did not address the\ suitability of OT type fuses for DC application. ihe vendor information specified AC ratings only but the fuse was'to be used in the 125 VDC cynrol circuit for the Diesel G6nerator ou'cput breaker.
, ( R7. tNo. 035) t
,AYuseControlProgramshouldensurethatadequate 4.1.2
/r information is included cn the drwings to support operations and maintenance. Deficiencies were rated on the N' design drawings for tfmatiove referenced ENCs. >These included: failure to indfeate fuse amarage on vbematic
- drawings; fallne'to change reference numbers, Mch specify
,,' type and ampcNge, on connction drwings; and .fa'41ure to
', revise Bill of Hateriabdrasthg which should s#cify new
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fuse type ar.d amperage.) (RI yto. 0?5) l I
4.1.3 AFuseControlPrograreshou[densurethatthedesign J
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- specified fuses are installed and properly mahhined. As a 0 -
sample,, diesel generator A output breaker was c% cked to d determine if proper fuses are installed. Design drawings show that the.2 fuses for protective relay 5ntshould be 2AK20. Actually installed was found an OT15 and a renewable fuse, ie' were unable to d'etermine the amperage. Design
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- OVERALL RESULTS AND' CONCLUSIONS (Continued)
" drawings show that the 2 fuses for metering should be 2AK20. Actually installed was found 2 renewable fuses, amperage. undetermined. Design drawings show that the single fuse for the control circuit for the diesel generator breaker should be a 2AK20. Design Calculation,.Z-EDS-E0671, says the renewable "REN-20" will be replaced with an OT10 because the "REN-20" does not-coordinate properly with the circuit breaker. What was actually installed was an FRN
-type fuse, could not see amperage rating. All five fuses in diesel generator A output. breaker-circuitry appear to be in variance with design bases. The diesel generator B output breaker was checked but all fuses were removed for work in progress.
No further survey of installed fuses was performed. At other sites renewable fuses are not allowed on site for the following reasons: (1) different amperage links are interchangeable and positive identification is impossible without disassembly, (2) due to mechanical connections, hot spots and unreliable operation are possible (3) possibility of these fuses being used in safety related circuits.
(RI No.035) 4.1.4 A Fuse Control Program should provide training on fuse application, fuse installation, and fuse maintenance. q Several design personnel and several electrical maintenance personnel were interviewed. All of the personnel interviewed said they had receibed no training on fuses or fuse control. They knew of no fuse training available. To their knowledge there are no procedures or guidelines dealing with fuses. Training was contacted, they said that, recently a 7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> course dealing with basic protective devices had been developed and some people had been trained. The adequacy of the training was not reviewed.
(RI No. 035) 4.2 Human Factors Concerns The inspection team has two concerns for the Emergency Diesel Generators System because of lack of Human Factors Engineering being applied to the EGS system.
(1) Possible damage may occur on the diesel generators before the operator can respond due to lack of human factors on the diesel gaugeboards and annunciator procedures.
j (2) During required surveillance or maintenance operation, j< if a Design Basis Accident were to occur, it may take the operator an inordinate amount of time to accomplish a restart on the diesel generator due to the lack of l
l human factors considerations.
1' 1 N- _ ___
OVERALL RESULTS AND CONCLLLSENJ (Continued)
- The Emergency Diesel Generators System Status Report listed seven individual problems of a Human Factors nature. SSR Problem Nos. l.,2, 4, 22, 28, 29 and 30. .The problems appeared to be addressed on an individual basis rather than addressing them as symptoms of a generic problem. With over 23% of the known problems being related, the inspection team looked at this area further and identified additional concerns.
4.2.1 Annunciator' lights on the diesO .nd generator panels should be grouped,ioutlined, or colore to distinguish between lights which indicate normal c.ondition, lights which indicate the diesel'is "not ready ~for auto start," and
' lights which indicate that a parameter is exceeded. For example, conditions which actuate the emergency shutdown circuit are not all outlined in blue lamenoid. " Crankcase pressure high" and " jacket. water temp. alarm" are outlined in blue lamenoid, but other annunciator lights.which cause this condition, "overspeed alarm" and "lo lube oil alarm" are not. (RI No. 079)'
4.2.2 Annunciator windows 92 (Diesel Generator A Auto Start Inoperable or Loss of Remote Control) and 93, (Diesel Generator B Auto Start Inoperable or Loss of Remote Control), each receive input signals from twenty. actuating p devices listed on six pages for each window in Annunciator Procedure H2ES, Rev. 14, which have to be checked before the operator can take action to correct the problem and make the diesel generator " ready for auto start" or " returned to Control Room" control. Considerable time is required to determine the problem before corrective action can be initiated. (RI No. 079) 4.2.3 Indicating Heters and gauges on diesel generator local panels and in Control Room do not indicate ' normal ranges" or " limits," such as Tech. Spec. limits or dangerous operating limits. (RI No. 079) 4.2.4 Three alarm lights on the diesel local annunciator panel are not addressed in the alarm response procedure. (RI No. 079) 4.2.5 Alarm lights on the local panel are numbered but do not correlate with the alarm response procedure. It takes considerable time to locate the appropriate response.
(RI No. 079)
i 5.0 DETAILED OBSERVATIONS _ RE00EST FOR'INFORMATIQ!i During' an evaluation, all potential concerns are documented on !
.Requcst 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 j the EASRTP team is having difficulty obtaining. {
l These RIs are considered drafts throughout^the entire evaluation- i until they become part of the final report. Responsible I organizations can accept the potential. concern as valid or they may l 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 1 members to better focus'the concern. RIs. developed during the system evaluation comprise this section of the report.
Attachment 2 of the report 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.
Approximately one week will be provided after the report is issued to provide time for departments- to address each RI for~ validity. A g revision to Attachment 2 will then be issued to reflect the status of RIs. All RIs not acknowledged at the end of this period will have an "Open" status. RIs are then transferred into the Restart Scope List tracking system for resolution and corrective action implementation.
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ATTACHMENT NO. I REQUEST FOR INFORMATION (RI)
RI NO: 035 SYSTEM CODE: EGS ISSUE DATE: 07-28-87
SUBJECT:
FUSE CONTROL DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR: RON LAHRENCE TEAM MEMBER: OURAISH DINAN l TEAM LEADER: THOMAS MARSHALL POTENTIAL CONCERN /00ESTION:
Potential for Emergency Diesel Generator failing to perform their required function because of inadequate fuse control.
. ECN's Ril50, R1128E and.R1060 were reviewed and the following problems were found: 1
- 1. None of the Design Bases Reports addressed fusing. The following questions were asked of Engineering:
- a. Has available Short Circuit Current and Interrupting ,
Capacity evaluated?. This was addressed on one of the ECNs, not for the other two.
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- h. Has time coordination with other protective devices evaluated? This was also addressed on one ECN but not the other two.
Et was addressed partially on one ECN but not on the other two.
- 2. Some of the Schematic drawings do not show the amperage rating for the fuse.
E-204 Sh. 1 Fuse to Metering and Synchronizing E-204 Sh. 2 Fuse to Metering and Synchronizing E-204 Sh. 65 FU3 and FU4 DC Air Start System E-204 Sh. 65 FUS and FU6 AC Air Start System E-204 Sh. 65 FU1 and FU2 Fault Shutdown Circuit E-204 Sh. 66 Fu.ses to' Alarm Lights and Annunciator E-204 Sh. 68 FU13, 14; 15 and 16 Generator Protective Relays Auto and Manual voltage Control, Electric Governor Control.
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- 3. The Bill of Material does specify the type and amperage of
' fuse but does not adequately , indicate where it is to be installed.
. Interviews with Maintenance. Personnel were conducted and the following concerns were expressed by Electrical Maintenance Personnel:
- 1. No training on Fuse Application or Fuse Installation.
- 2. I' adequate labeling of fuses.
- 3. lesbility'to determine fuse type and amperage from drawing.
- 4. The lack of guidelines or procedures addressing fuses.
- 5. If a fuse was damaged or had been previously removed, it might be impossible for maintenance to determine the typc or even amperage of the required fuse because of inadequate information on the drawings.
- Observation by Evaluation Team
- 1. Inadequate labeling of fuses in EGS circuits.
- 2. Inadequate controls on. storage of fuses. Replaceable element fuses were stored along with other fuses.
- 3. Replaceable element fuses were observed in the control j circuit for Diesel Generator A output breaker. Electrical Maintenance Engineer stated that, Renewable Fuses are not used in Safety Related Equipment.
- 4. Of 5 fuses in the Diesel Generator A output breaker compartment, all 5 appear to be not in accordance with the drawings.
.- NCR No. 5535 contains details of a fuse failure on' Emergency Diesel Generator "B" because of inadequate fuse control.
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ATTACHMENT NO. I REQUEST FOR INFORMATION (RI)
RI NO: 079 SYSTEM CODE: EGS ISSUE DATE: 07-29-87
SUBJECT:
HUMAN FACTORS DEPARTMENT: OPERATIONS COORDINATOR: RICH MACIAS TEAM MEMBER: MIKE FOSTER TEAM LEADER: TOMMY MARSHALL POTENTIAL CONCERN /00ESTION:
Possible damage may occur on the diesel generators before the operator may be able to respond due to the lack of human factors on the. diesel gaugeboards and annunciator procedures. During a diesel generator surveillance with a designed basis accident it may take the operator a great. deal of time for a restart of the concerned diesel generator due to the lack of human factors considerations.
. Lighting on the north side of G-886B diesel generator room is inadequate.
- There is no reflash capability for Control Room annunciators for the diesel generators; H2ES windows 1, 18, 92, and 93.
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. Annunciator Procedure H2ES, Rev. 14, index has incorrect window names for windows 92 and 93, although they are correct in the procedure.
. Control Room annunciator procedures for the diesel generators do not reference local' panel annunciator procedures.
. Annunciator windows 92 (Diesel Generator A Auto Start Inoperable or Loss of Remote Control) and 93, (Diesel Generator B Auto Start Inoperable or Loss of Remote Control), each receive input signals from twenty actuating devices listed on six pages for each window in Annunciator Procedure H2ES, Rev. 14, which have to be checked before the operator can take action to correct the problem and make the diesel generator " ready for auto start" or " returned to Control Room" control. Considerable time is required to determine the problem before corrective action can be initiated. (RI No. 079)
. Meters on diesel generator panels H7J272 and H7J273 and generator panels H2DGA and H2DGB are not color coded for procedure " Limits and Precautions" and " Process Standards" setpoints.
. Pumps are not labeled with noun equipment names even though A.31, Diesel Generators System, Rev. 28, refers to noun names.
. Annunciator lights on the diesel and generator panels should be grouped, outlined, or colored to distinguish between lights which indicate normal condition, lights which indicate the diesel is "not ready for auto start," and lights which indicate that a parameter is exceeded. For example, conditions which actuate the emergency shutdown circuit are not all outlined in blue lamenoid. " Crankcase pressure high" and " jacket water temp. alarm" are outlined in blue lamenoid, but other annunciator lights which cause this condition, "overspeed alarm" and "lo lube oil alarm" are not. (RI No. 079)
. Engraved Nameplates on panels H7J272 and H7J273 do nct match the names of alarms in' Annunciator Procedures H2DGA, Rev. O and H2DGB, Rev. O.
. There is a lack of standardization in diesel fuel oil day tank level indication.
- 1) T-893A and T-893B sightglasses have no markings.
- 2) LI-88693 and LI-88694 are marked in fractional increments of !
thirty-seconds, sixteenths, eighths, and halves.
- 3) Process Standards, AP.159, Rev. 7 and Annunciator Procedures H2DGA, Rev. O and H2DGB, Rev. O, use inches.
Diesel Generator System, A.31, Rev. 28, uses inches and I
4) percentage.
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- 5) Technical Specification Amendment 147, Section 3.7.IF, will require percentage or gallons.
. Not all alarm lights on local annunciator panels are included in alarm response procedures H2DGA, Rev. O and H2DGB, Rev. O. During STP-1012 on G-886B alarm light " Comb. Air Press Lo" annunciated and the operator had no idea what to do because the alarm was not listed in Annunciator Procedure H2DGB, Rev. O.
. Local Panel Annunciator Procedure H2DGA, Rev. O and H2DGB, Rev. O are not easily referenced. During STP-1012 on G-886B the alarm light for " motor driven fuel pump failure" annunciated and it was several minutes before the operator could locate the alarm in the annunciator procedure to determine a response, even though lights on diesel panel H7J273 are newly numbered, t
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6.0 ATTACHMENTS 6.1 List of Documents Reviewed 6.2 Status of RIs
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' :L ATTACHMENT 1 6.1 LIST OF DOCUMENTS REVIEHED USAR Section 8 System Status Report for EGS, Rev. 1 Rancho Seco Technical Specification Draft Copy of System Design Basis, NEPH 5422 Quality Control Instruction QCI-12, Rev. 3 Master ECN List Installed Parts Equivalency Evaluation Report, AP.91 Field Problem Report, RSAP-0305, Rev. 0 :
Maintenance Test Procedure, RSAP-0807, Rev. O Rancho Seco Annunciator Procedures H2ES, H2DGA, H2DGB Design Guide 5204.59 >
Construction Specification 5304.11C Regulatory Guide 1.137, October 1979, Fuel Oil System for Standby Diesel Generator AP.44,. Rev.11, Plant Modifications Design Bases Reports for ECN R-1060, R-il50, and R-1128E ECN Package for ECNs R-1060, R-il50, and R-il28E -
RSAP-0803, Rev.1 & 2 - Hork Request Reviewed the Hork Request Summary for Trending, both Nucleis and MIMS IE Notice 85-91 Licensee Event Report 87-08 I Occurrence Description Report 46 SP.206.01A Biannual "A" Emergettcy. Diesel Eng. Insp.
Emergency Operating Procedures AP.23.10, Equipment Maintenance and Operating Standards AP.91, Installed Parts Equivalency Evaluation Report STP.1012. Emergency Diesel Generator (G-886B) Post Modification Test Rancho Seco Daily Plant Outage Schedule Field Problem Report #14, ECN R-1898 Field Problem Report #15, ECN R-1150 A.31B (Proposed), TDI Diesel Generator System OP C.13A (Proposed), Remote Shutdown AP.159, Rev. 7, Diesel Generator and Diesel Fuel Oil Systems Technical Specifications Manual, Amendment 147 SP.206.05, Rev. 3, Standby Diesel Generator and Diesel Driven Fire Pump Fuel Oil Quarterly Testing AP.306, Rev. 2, Chemistry Manual ECN'R-1108. R-0770A/B, R-0890, R-0415A/B, R-1000 STP-970, STP-895, STP-1062 EM.126B ;
P&ID M-583, Sh 1, Rev. 12 RSAP-0807, Rev. O, Maintenance Test Program Emergency Diesel Generator System Test Matrix a
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' ATTACHMENT 1 (Continued)
J '.
Surveillance Procedures. .
-SP.206.01B, Rev. 3, Biannual Emergency. Diesel Generator Inspection-SP.206.03A, Rev. 16, DG G-886A,.-Synchronization Surveillance.
. Test SP.206'.03B, Rev. -14, DG G-886B, Synchronization Surveillance H '
Test' l' SP.206.01A, Rev. 3, Biannual "A" Emergency Diesel Engine Inspection' SP.206.02A, Rev. 4, Refueling Interval Diesel. Generator "A"
'SFAS Start Test SP.319A,- Draf t Copy Refueling Interval' DG (G-886A)- SFAS and LOOP Loading Scheme Surveillance Test Electrica1' Maintenance Procedures EM.177A, Rev. 8, function Test of Nuclear Service Bus "A" and.A2 Unloading Scheme ,
EM.126A, Rev. 1, Refueling". Testing and Maintenance of Diesel Generator A EM.164, Rev. 5, Testing of Molded Case Circuit Breakers-EM.144, Rev. 9, Testing of Protective and Control Relays EH.168, Rev. 5, Electrical Maintenar;.ce Inspection of Emergency Diesel Generators.
MAP-0004, Rev. O, Control of Onsite Contractor.and Vendor Personnel MAP-0006, Rev. O, Work Request \ Planning
' Operation Proceduros A.32A, Original Issue (Draft), Diesel Fuel Oil System-A.31,-Rev. 28, Diesel Generator System Casualty Procedures C 13A,-Rev. 6, Hot Shutdown from Safe Shutdown Panel Hith a' M re it, Dantrol Room C.180 (Proposed), Diesel Generator A(B) Fail to Start C.181 (Proposed), Diesel Generator A2(B2) Fail to Start !
C.182 (Proposed), Diesel Generator A(B) Fail to Load 1 C.183 (Proposed), Diesel Generator A2(B2) Fail to Load '
l C.184 (Proposed), Diesel Generator A(B) Fail to Stop C.185 (Proposed)', Diesel Generator A2(B2) Fail to Stop Calculations- -l J
Z-EGS-E0658, Rev.1, Size Verification of Diesel . Generators GEA, GEB, GEA2, GEB2 Z-EDS-120, Rev. 2, Short Circuit Study of Aux Power System Z-EDS-E0671, Rev.1, Control Circuit Fuse and Breaker Coordination for Switchgears S4A2, S4A and Load Center a
t ATTACHMENT 1 (Continued)
Drawinas E-104, Sh 2, Rev. 11 - One Line Diagram, 4160V System P&ID M-582 Diesel Oil Systems E-104, Sh 3, Rev. 12 - One Line Diagram, 4160V System P&ID M-583 Emergency Diesel Generator E-204, Sh 1 .Rev. 23 - Elementary Diagram, Diesel Generator System E-204, Sh 2, Rev.18 - Elementary Diagram, , Diesel Generator.
System ;
E-204, Series Drawings for Diesel Engine "A" and "B" Control E-208 Series Drawings for Nuclear Service Bus Loading i Hark Reauest ,
Work Request Nos. P65830, 125891, 116927, 133702, 134215, 125787, 128900, 117439, 120433, 120434 l
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. ATTACHMENT 2 6.2 STATUS OF'RIs RI NUMBER SUBJECT ' STATUS RSL NUMBER 035.. Fuse Control' Acknowledged RSL-RI-035 079 . Human Factors Acknowledged RSL-RI-079 v.
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.A EXPANDED AUGMENTED SYSTEM REVIEW AND TEST PROGRAM (EXPANDED ASRTP)-
EVALUATION OF THE MAIN STEAM SYSTEM SUBMITTED BY: DATE: 9-l'87 FRANK STOCK TEAM LEADER CONCURRENCE: wi u%t DATE: bl-77 f
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VID HUMENANSKY EXPANDED ASRTP PROGRAM ANAGER CONCURRENCE: DATE:
/ '
/ BOB CROLEY.
DIRECTOR, NUCLEAR TECHNI AL SERVICES I
WP3425B/D-0183B
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i TABLE OF CONTENTS-
.Page Number i
10 INTRODUCTION- 3 4
- 2.0 PURPOSE 5
3.0 SCOPE 4.0 OVERALL RESULTS AND CONCLUSIONS '6 9
5.0 DETAILED'0 OBSERVATIONS - REQUESTS FOR INFORMATION.
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- f. 0 ATTACHMENTS .10
.6.1. List of Reviewed Documents 6.2 Status of RIs g
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" j EXPANDED AUGMENTED SYSTEM REVIEW AND TEST PROGRAM d
EVALUATION OF THE MAIN STEAM SYSTEM l l
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1.0 INTRODUCTION
The Rancho Seco Expanded Augmented System Review and Test Program 4 2
[ASRTP]. evaluation effort involves an assessment of the l effectiveness of the System Review and Test Program [SRTP] and an analysis of the adequacy of ongoing programs to ensure that systems j will continue to function properly after restart. The Expanded ASRTF is a detailed system by system review of the SRTP as implemented on 33 selected systems and an in-depth review of the engineering, modification, maintenance, operations, surveillance, inservice testing, and quality programs. It also. conducts a review, on a sampling basis, af many of the numerous ongoing verification '
't...o review programs at Rancho Seco.
Six multi-disciplined teams composed of knowledgeable and .l experienced personnel are tasked with performing the Expanded 4 I
ASRTP. Each multi-disciplined team consists of dedicated personnel with appropriate backgrounds to evaluate the operations, 4 maintenance, engineering, and design functional areas.
f Independence, perspective, and industry standards provided by team members with consultants, architect engineer and vendor backgrounds arejoinedwiththespecificplantknowigdgeofSMUDteammembers.
.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 enhance the evaluation. Each team prepares a final report for each completed selected system evaluated. This report is for the main steam system, 4
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3 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 plantioperation after restart.
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3.0 SCOPE p
To accomplish the first objective, the steam plant team evaluated !
the main steam system to determine whether:
- 1. The system was capable of performing the safety '
functions required by its design bases. I
- 2. Testing was adequate to demonstrate that the system 4 would perform all of the' safety functions required. j i
- 3. System maintenance (with emphasis on pumps and valves) {
was adequate to ensure system operability under i postulated accident conditions.
- 4. Operator and maintenance tbchnician 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 pr,oper system operations under normal and accident conditions.
To accomplish the second objective, the steam plant team reviewed the programs as implemented for the main steam 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 steam plant team reviewed a number of documents in preparation for and during the Expanded ASRTP evaluation of the main steam system. This list of documents is found in Attachment 1.
The primary source of leads for the team were the problems identified in the Main Steam 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 main steam system included a review of pertinent portions of support systems that must be functional in order for the main steam system to meet its design objectives.
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~ j 4.0 OVERALL RESULTS AND CONCLUSIONS l 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. Section 5.0 provides detailed findings by providing the Request for-Information i (RI) forms that are used by the Expanded ASRTP teams to identify potential concerns during the evaluation. The numbers in brackets
. af ter each individual summary refer to the corresponding RIs in i 1
Section 5.0.
i 4.1 Design and ' Engineering I
4.1.1 A system walkdown revealed that the current atmospheric dump valve (ADV) and turbine bypass valve (TBV) configurations did not adequately provide for thermal expansion and valve' movement during operation. Specific deficiencies were noted with respect to: l i
. inadequate lengths of flexible hosing f '
. possible interference with conduit and nearby structures
. inaccessibility to a handwheel operator on an ADV ;
instrument air tubing which will move, and possibly-
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fracture, during valve operation. [RI 19] [RI 97]
Since these deficiencies pertain to a number of ADV and TBV's in the plant, the potentigl for loss of operability of one or more of these valves is apparent.
4.1.2 The USAR, system design bases document, and the applicable surveillance procedure do not utilize the same acceptance criteria with respect to the main steam line break analysis and closure time for turbine-throttle stop valves. The surveillance procedure utilizes a value approximately 15 times greater than that contained in the USAR and therefore does not verify that the acceptance criteria contained in ,
the USAR with respect to the accident analysis are valid.
This inconsistency was identified in NCR-5576, and later voided by memo, however the explanations provided did not address increased OTSG blowdown. It appears as though this analysis should be reevaluated. [RI 15]
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OVERALL RESULTS AND CONCLUSIONS (Continued)
'4.1.3 A review of the System Design Basis, Revision 1 (Draft),
indicated that main steam is no longer addressed as a secondary source of steam to the gland seal steam (GSS) system via the auxiliary steam header. .
A review of the related P& ids, line designation sheets, and the Master Equipment List (MEL) identified the fact that main steam, at 900 psig, can enter the auxiliary steam header if MSS-035'is open. Since there is no pressure relief valve or no pressure reducing valve downstream of the MSS valve, this could result in a pipe rupture or component failure resulting in a loss of the ASC header and valves,.
and plant auxiliaries. [RI 46]
4.1.4 The System Status Report identifies four (4) problems related to the performance and reliability of the MSSV '
acoustic monitors. An inspection of the system design, installation, and test requirements indicates that the poor reliability and performance may be the result of;
. excessive lengths of cable which are unsupported or not run in conduit for all 18 valves
. bare signal cable strapped to the relief pipe for g 3 valves
. calibration procedures which do not provide for performance evaluation The performance and reliability of the MSSV acoustic monitors is necessary as a means for the monitoring of radioactive releases in accordance with 10CFR50, Appendix A, Criterion 64. [RI 10]
4.1.5 In response to SSR Problem #10, the installation of the backup instrument air supply was evaluated. The latest ,
calculations for the backup instrument air supply contained 1 inaccurate assumptions regarding air consumption and bottle depletion rates. In addition, the corresponding special test procedure which-is to be utilized for verifying the calculation data and alarm setpoints references an earlier revision of those calculations.
This same surveillance test procedure is also intended to verify the identified bottle depletion rates and actual air consumption rates, however, the test is performed in a cold condition without steam flow. The amount of air consumed by the valve positioner will actually be greater when hot and with steam flow present since the valve will require more force to stroke. Therefore the special test procedure is not adequate for verifying bottle depletion rates and air consumption,[without performed. RI 37] [RI some 47] additional testing being l t .
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'-* , '0VERALL RESULTS AND CONCLUSIONS (Continued)
~4.2 ' Programmatic 4.2.1 Several instances were no etdi
. n which equipment / components had not: been installed'in accordance with vendor
' recommendations / instructions. .,[RI.10):[RI 66] [RI 14].
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5.0 DETAILED OBSERVATIONS - REQUEST 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. J These RIs are considered drafts-throughout the entire evaluation i
until they become part of the final 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 report.
Attachment 2 of the report 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 i RI An RI would also be closed if requested information das provided. All other RIs are open. Acknowledged RIs are ) pen RIs ,
that have been accepted as valid by the responsible organization. l Approximately one week will be provided after the report is issued to provide time for departments to address each RI for validity. A il j
revision to Attachment 2 will then be issued to reflect the status of RIs. All RIs not acknowledged at thg end of this period will l have an "Open" status. RIs are then transferred into the Restart j Scope. List tracking system for resolution and corrective action implementation.
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REQUEST FOR INFORMATION (RI)
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RI NO: 010 SYSTEM CODE: MSS ISSUE DATE: 07-20-87
SUBJECT:
MSSV ACOUSTIC MCuTTORS DEPARTMENTS SYSTEM ENGINEERING COORDINATOR: JOHN IITNER
-TEAM MEMBER: T. LOVETT TEAM LEADER: F. STOCK E. ORTEGA POTENTIAL CONCERN /0VESTION:
The design and installation of the acoustic monitors does not provide a reliable means for complying with requirements of 10 CFR 50, Appendix A, Criterion 64, with respect to " monitoring of radioactive releases." The SSR identified 4 problems relating to the acoustic monitors. The following items have'been identified with respect to the above and capabilities for monitoring of fsite dose releases.
a) Visual inspection of the signal cable installation
- 1. Bare signal cable strapped to relief pipe could affect the transmission of the signal itself due to temperature and grounding considerations. 5
- 2. . Excessive lengths of cable are s coiled and left unsupported. This could introduce unnecessary signals due to pipe or wind motion.
- 3. The existing installation, in about half the cases, does_
not meet the manufacturer's recommendation stated in TEC Technical Manual (30120-0M-01), Section 2, " Sensor Cabling." This requires that these cables be supported or run in conduit.
b) Review of related calibration and test procedures, did not provide for specified calibration frequencies or methods to provide for performance evaluation.
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- b REQUEST FOR INFORMATION (RI) y RI NO: 014 SYSTEM CODE: MSS ISSUE DATE: 07-21-87 j
SUBJECT:
CONTINUITY OF OPERATIONS BETWEEN 0.P.. A.6. A.46 AND A.53 PROCEDURES DEFARTMENT: OPERATIONS COORDINATOR: R. MACIAS TEAM MEMBER: HARLAN COOMES TEAM LEADER: FRANK STOCK PROCEDURES REFERENCED:
O.P. A.6, Steam Generator Secondary Side System, Revision 26 '
Main Turbine System, Revision 20 l 0.P. A.46 L
0.P. A.53 Extraction Steam, Reheater and Feedwater Drain System, Revision 20 ,
Concern The MSR main steam coil could be subjected to uneven heatup and possible tube bundle failure. The procedures and valve line ups used in the present configuration are not in conformance with Westinghouse recommendations for MSR main steam coil isolation, evacuation and startup. j The Westinghouse recommendations in I.L. 1370-1482, Rev.1, state:
- 1) " Positive isolation of the high pressure tube bundles from heating steam must be assured until the turbine generator reaches the specified load for reheater activation. A leak off valve must be located downstream from the main steam shut-off ,
valve to prevent leakage through the control valve during the startup operation."
- 2) "It is necessary to evaluate the non-condensibles from the tube bundles prior to the admission of steam in order to warm the tubes evenly."
- 3) " Controlled admission of heating steam to the high pressure tube bundle is required in order to avoid excessive thermal shock."
- 4) " Venting of .2% of the heating steam entering each tube bundle must be vented through the vent condenser over all operating conditions." ..
" Failure to provide.this positive isolation system can result in premature and uneven heating of the tubes with subsequent tube bowing and i tube failures." l 1
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4 RI NO: 014 (Continued)
The valve line up in A.53 is used as a: prerequisite for performing
~Section 4.2 of A.46. The valve line ups in A.53 and A.6 conflict with
.the desired initial position of valves manipulated in A.46, Section 4.2. l The MSR main steam coil purge block valves,: main steam coil regulator block valves and the main steam coil- 1" purge . valves .are required to be open per the normal valve line up. A.46 directs' opening of these valves
'in the procedure. This will result in steam being admitted to the MSR
' main steam coil when it is not desired. The condenser leakoff lines are closed in the valve line'up and A.46 directs closing these valves. The valves should be-open during heat up to prevent steam inleakage to the ;
main steam coil. The main steam coil vents to condenser are closed and should be open to provide tube bundle evacuation during heat up.
Procedures A.46 and A.53 do not direct manipulation of the main steam i coil; vents. ]
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REQUEST FOR INFORMATION (RI)
RI NO: 015 SYSTEM CODE: MSS ISSUE DATE: 07-22-87
SUBJECT:
MAIN TURBINE THROTTLE /STOP VALVES STROKE TIMES DEPARTMENT: SYSTEM ENGIN'EERING . COORDINATOR: J. ITTNER TIM LOVETT TEAM LEADER: FRANK STOCK TEAM MEMBER:
POTENTIAL CONCERN /00ESTION:
The USAR main steam line break analysis may be a non-conservative assumption that does not reflect the actual closing time of the turbine throttle stop valves. The assumptions used in the USAR steam line failure analysis have not been verified by the surveillance procedure, nor is the USAR analysis consistent with the surveillance procedure and the System Design Bases Manual, NEPM 5444.
Section 14.2.2.1 of the USAR steam line f ailure analysis assumes that turbine throttle stop valves close within 0.5 seconds after receiving a trip signal. NEPM 5444, System Design Basis, page 9, states " Turbine stop valves have minimum and maximum closing time of 0.20 and 0.30 seconds respectively. The stop valve closing times used in the steam line break analysis is 0.25 seconds." Section 10.3.1 of the USAR states that 0.25 seconds is the closing time used in the steam line break accident analysis and also states that the closing g time during normal operation is 0.20 seconds minimum to 0.30 seconds maximum. The surveillance procedure SP.210.03C, " Turbine Throttle Stop Valves Fail Test," lists as an acceptance criteria, "The turbine throttle stop valves shall move f rom open to closed in three seconds or less." The discrepancy between the value used in the USAR accident analysis and the SP acceptance criteria was addressed in NCR 55576. This NCR was later voided by memo SRT 86-154. The memo states that the three second acceptance criteria "will be maintained as a conservative limit and check." The explanation given is " Excessive Flow through the High Pressure steam chest will peak at a value equivalent to over 400% of full Reactor Power. Since the closure time of the Turbine Throttle Stop Valves is affected by the Steam Flow Rate, this will" have the effect of reducing the Throttle Stop Valve closing time." .. The Turbine Stop i Valve closing time has, in fact, not been precisely measured," and "The existing surveillance does not model the USAR Analysis and that we will l
not be able to model the USAR Analysis via surveillance."
The explanations for voiding the NCR 55576 are not supported by ,
l calculation 2-MSS-M0829 that shows that 6 seconds after the break, steam flow has decreased to 112% of rated flow. The 400% of full reactor power
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steam flow noted in the memo is used as a basis for the assumption that the closing time of the Turbine Throttle Stop Valves will be reduced by steam flow through the H.P. steam chest.
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RI NO: 015- (Continued)
The reasons for voiding NCR 55576 do not address the possible consequences of extending the. blowdown time of'both OTSGs. It appears that a reevaluation.of the discrepancies in the USAR, surveillance procedure and design bases for the throttle stop valve closing times is
- necessary.
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-REQUEST FOR INFORMATION (RI)
RI NO: _; 019 SYSTEM CODE: MSS ISSUE DATE: 07-22-87
SUBJECT:
. ADV & TBV INTERFERENCE CAUSED BY THERMAL EXPANSION NUCLEAR ENGINEERING- COORDINATOR: RON LAWRENCE DEPARTMENT:
WAYNE ADACHI TEAM LEADER: FRANK STOCK TEAM MEMBER:
POTENTIAL-CONCERN /00ESTION:
During operation the valve movements could either damage the valve (s) or valve positioner or-the instrument air tubing and'could eventually render the. valve inoperable.
The existing MSS configuration does not provide adequate movement clearances around the atmospheric dump valves (ADV) and turbine by-pass valves (TBV) caused by operational tnermal expansion, design loads, flow loads and valve operation. ,
For the ADVs, the following total movements occur f rom calculation 2-MSS-M2055, Rev. 0:
Inches 1
AY AZ
> Valve No. aX PV-20562A - 4.275 + .05 - 3.48 PV-205628 - 4.28 .039 - 3.487 PV-20562C - 4.297 .038 - 3.431 PV-20571A - .943 + .237 - 1.575 PV-20571B - .939 .087 - 1.594 PV-20571C - .939 .053 -
1.627 For the TBVs the following total movements occur from calculation Z-MSS-M0430 Rev. 3, M0431 Rev. 4:
Inches PV-20561 + 1.071 + .329 + .606 )
PV-20563 + .792 & .498 + .655 J PV-20564 + 1.452 + .231 + .889 PV-20566 -
. 97 + .653 ~ + .997 l o PV-20571A in train A has movements such that there is valve positioner interference with the newly installed conduit 710013 and I 710084 and conduit support #169.
PV-20562A in train B has 4.28" movements in the north direction and 1 o '
has possible positioner interference with conduit support #182 when the valve opens and creates vibration of the valve.
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'All of the valves have
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installed ?? xible stainless, steel 3pj g .: tubing 4 Some' of the. ficlible tubing arefinstalled such thitj the c ( .L 04 '
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I. . valve 349errents can stretch and damage t.E> tubing. /V j .m ,
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,. 4 The nw' fler%1e M!rqwnt air tubing is mounted ;nsidei [, '
i/a. support and exist 0 4t rira ht'iingles around a corner of thy ;2 !
fA'+d , ' ;J44 -
,6 di . 3 j0J 9The W,bpport. . h. braided' PJL20561. tehing and the corner are about 1/?'{ bart. '
^ '" ' /Withapproximatelyoneidhmovementinthesouthdirectipp,[the 3 )
D gf, br'dded tubin will cost /et sthe support corner and the braided 'J r g -
tubing could ecome d waged , The two instrument air tubings to the a
.4. ' 4 s t valve near 3he positirg ceross/over. The two tubes are close , ' (. ;
enough that they coul>fant.qcotact during pipe vibratinrland risult~ <
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V3}- in tube damage. y'Q \,f" M A 'I g.g - ?
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5 ,, . PV-20564. The two new instramer.t air flexible tubings to_the TC/
are' installed with no tubing slack to account for the,1.452' .
c- movendt.t. in' thesduth direction. The resulting c4vem nt coifd ,
damage 16e' flexible tubing. [ ' ,:
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' REQUEST FOR-INFORMATION (RI) !
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037' MSS ISSUE DATE: 07-24-87 l 3
RI h0: SYSTEM CODE:
SUBJECT:
CALCULATION AND VERIFICATION OF BACKUP INSTRUMENT AIR SUPPLY NUCLEAR ENGINEERING COORDINATOR: RON LAWRENCE DEPARTMENT:
WAYNE ADACHI TEAM LEADER: FRANK STOCK TEAM MEMBER:
I POTENTIAL CONCERH/0VESTION:
i The calculations and surveillance procedures utilized for design and test {
I of backup instrument air supply do not contain proper assumptions and are not performed under conditions which will verify operability in f accordance with specified requirements.
The Design Basis Report for ECN-A5743, Rev. 6 specifies that a procedure will be performed to test the backup instrument air system gas bottle supply in case of a loss of the normal instrument air supply. The .
following is to be performed:
"In addition to pressure integrity test, an STP will be written to verify the data used in the calculation No.
Z-IAS-M2094 for air consumption / bottle depletion rate when {
the ADV's are being operated. Valve stroke time under system pressure should also be verified sas'part of this '
test. The test should also verify the IDADS alarm upon degradation of the system pressure,"
STP.744, Rev. O, addresses the first two items, however the IDADs alarm setpoint is not defined. Reference is made to calculation Z-IAS-M2084, Rev. O, which indicates pressure alarm setpoints for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (1200 psig) and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (750 psig) to 90. STP.774 indicates a single alarm in each train (P-1775 for train A and P-1774 for train B). The calculation
. indicates'that only one alarm is required, preferably the "2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to go" pressure setpoint. The test personnel have no way of determining which value to use from this STP. Subsequently Rev. 2 of calculation 2-IAS-M2084 has been calculated for the "2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to go" pressure of 850 psig. The current STP'does not reflect this change since it only references Rev. O of the calculation.
A similar type of problem exists for the Appendix "R" gas bottle where the P-1780 alarm setpoint is not identified in the procedure and there is a difference in values from the referenced Rev. O and the subsequent Rev. 2 of the calculation.
'R1 NO: 037 (Continued)'
In calculation 2-IAS-M2084,Rev. 2,. sheet'7A, the volume-of the valve
' modulation was calculated and the total volume accounted for only a-single valve modulation'in the train. Assumption D'on page 3 incorrectly states that only'onelADV is in service at a time. In reality it is possible to have three valves per train in service at the same time. The calculation should be revised to ensure no invalid results have been utilized.
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REQUEST FOR INFORMATIO (RI)- ,
RI NO: 046- SYSTEM CODE: MSS ISSUE DATE: 07-27-87
SUBJECT:
POTENTIAL FOR HIGH MSS PRESSURE TO ENTER-LOW PRESSURE AUXILIARY STEAM HEADER DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR: RON LAWRENCE TEAM MEMBER: WAYNE ADACHI TEAM LEADER: FRANK STOCK i
REF'ERENCES
- 1) P&ID M-530, sheet 3, Rev. 17 2)' SMUD Rancho Seco Nuclear Generating Station, Systems Training
' Manual-Main Turbine-Chapter 16 and 16 c.
- 3) Main steam system design bases, Revision 1 (advance copy)
- 4) MEL for GSC and ASC
- 5) Line designation sheet no. 158, 162, 298
- 6) Piping Design Specification Rev. 9
- 7) P&ID M-537, sheet 3, Rev.13 POTENTIAL CONCERN /0VESTION:
The low pressure ASC header could be exposed to main steam pressures of f 900 psig, which is beyond the ASC component ratings. This can result in-a pipe rupture or component failure causing ga loss of plant auxiliaries due to loss of the ASC header. Additional piping and component damage could also' result from the energy release from the rupture or failure.
The main steam at 900 psig (from Ref. 5) is normally supplied to the ASC through a pressure reducing Valve PV-36014A and/or PV-360148 (see Ref. 7) at a maximum of 250 psig. An alternate steam source, as identified in Ref. 2 and shown in Ref. 1, is through valve MSS-035 and line 30125-4"-EAI to the gland seal steam. MSS-035 is shown as normally open. This is an error since the plant configuration is normally closed. However, with MSS-035 open, the 900 psig main steam can potentially enter the ASC header since there is no pressure reducing i
valve downstream of valve NSS-035.
I The Ref. 4 MEL sheets for the ASC header were reviewed for maximum )
pressure that would occur in the ASC valves. These valves are:
VALVE MAX. PRESSURE (PSIG) )
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ASC-470 500 1 500 I ASC-587 ASC-603 500 )
ASC-601 500
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RI:' NO: '046:.(Continued)
Other, surrounding val'ves did.not have a maximum pressure. listed on the ME L, ..
It is possible that with MSS-035'open, that the above ASC valves and the ASC header will be exposed to the 900 psig main steam pressure, which is
- above the maximum pressure stated in the MEL.
The Ref.i 3 system design bases does not discuss this line.
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REQUEST FOR INFORMATION (RI)
.RI NO: 047 SYSTEM CODE: _,. MSS /IAS ISSUE DATE: 07-30-87
SUBJECT:
DESIGN BASIS TESTING OF BOTTLED AIR BACKUP FOR ADVs DEPARTMENT: SYSTEMS ENGINEERING COORDINATOR: JOHN ITTNER TEAM MEMBER: TIM LOVETT- TEAM LEADER: FRANK STOCK POTENTIAL CONCERN'/00ESTION CONCERN: STP-774, Rev. 0 will verify that the backup bottle system has the capacity to meet the design basis requirements for the valves in the .
cold condition with no steam flow, however it will not verify adequate capacity for the valves in actual opdrating condition. _Since the valve requires more force to stroke when it is hot with steam flow present, the total amount of air consumed by the positioner will be greater.
Therefore, STP-774 alone is inadequate to verify backup bottle system capacity.
The Design Basis for the bottled air backup system for Atmospheric Dump l Valves.(ADVs) and Turbine Bypass Valves (TBVs) includes a requirement that the bottled air backup systems provide the capability to " operate and control" the valves for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, for the EFIC or QA Class 1 backup $
system, and 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> for the Appendix R or QA class 2 system. STP-774, Rev 0 is intended to verify that the backup httle system has sufficient capacity to meet the desicn basis. However, STP-774, Rev. O is performed with ADVs and TBVs cold with no steam flow. ,
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- REQUEST FOR INFORMATION (RI)
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RI NO: 066 SYSTEM CODE: MSS ISSUE DATE: 07-29-87
SUBJECT:
. SNU88ER INSTALLATIONS DEPARTMENT': MAINTENANCE COORDINATOR: J. DARKE HARLAN COOMES TEAM LEADER: FRANK STOCK ,
TEAM MEMBER: l I
j POTENTIAL' CONCERN /00ESTION ,
REFERENCES:
o M.21.08A ITT/Grinnell Hydraulic Snubber Vendor Manual ,
o M.40, Rev. 4 Maintenance Proceaure " Removal / Repair / Installation of Hydraulic Shock and Sway Arrestors" o Technical Specification 3.12 and 4.14 M-486 SHT. 7-64 Rev. 2 M-486 SHT. 7-67 Rev. 2 H-486 SHT. 7-70 Rev. 2
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POTENTIAL CONCERN:
A variety of snubber mounting and pinning methods / techniques are utilized within the plant without specific criteria identified. Two of the MSS technical specification snubbers (75W-30800-12 and 75W-30800-14) have been installed in a manner which does not conform to vendor recommendations contained'in technical manual M21.08A.
It was also observed that some seismic I snubbers in the main steam system do not contain f astening or locking devices on those which have attaching pins of "all-thread" or " load studs."
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f "' REQUEST FOR INFORMATION (RI) !
RI NO: 097 SYSTEM CODE: MSS ISSUE DATE: 07-31-87
SUBJECT:
' BACKUP INSTRUMENT AIR FOR THE ATMOSPHERIC DUMP VALVES n --
DEPARTMENT: SYSTEM ENGINEERING COORDINATOR: J. ITTNER TEAM MEMBER: R. FISH TEAM LEADER: FRANK STOCK M. HORHOTA i
POTENTIAL CONCERN /00ESTION Backup instrument air tubing for the ADVs could fracture as a result of valve movement, leaving backup instrument air inoperable for the valves.
Inconsistencies exist in the installation of the backup instrument air
. tubing.
DETAILS:
During a system walkdown and review the following discrepancies were noted for the ADV backup instrument air tubing:
o Various horizontal and vertical tubing runs between the ADVs and I the control panel were found to have axial movement through their clamps, while other tubing rugs were clamped tight, allowing no movement.
o~ Tube clamps along various runs were found to be the incorrect' size allowing movement of the tube in the clamp (3 axis).
o Tubing clamps were found with no jam nut installed, jam nuts.
installed loose, or jam nuts installed with the clamp loose.
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' ' rt 6.0 ATTACHMENTS 6.1 List of Reviewed Documents !
6.2 Status of RIs ll 8
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LIST OF REVIEWE0 DOCUMENTS USAR System Status Report Technical Specifications Master Equipment List (MEL)
MIMS Nucleis Process Standards System Design Bases Overall Plant Operating Procedures System Operating Procedures Emergency Operating Procedures Annunciator Procedures Casualty Procedures g Special Test Procedures Surveillance Procedures Station Manual QA Manual l
P& ids l System Training Manual (s)
System Lesson Plan Maintenance Manual Line Designation Sheets 158,162, and 248 -
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- LIST OF REVIEWED DOCUMENTS (Continued)
ECNs:
A-5415C, A-5415Z, A-5415AE, A-5743, R-0043, R-03578, R-0518, R-0828, R-0861, R-1003, R-1367, A-3683-H, A5106, A5152, A5546, R-0859, R-0968V, R-09681, R-0968Z, R-0968ZZ NCRs:
5-5577 Rev. 3, S-5576, S-5885, S-5512 Rev.1, .S-5507 Rev.1, and -
S-5328 Rev. 2 WR:
83890, 118092 through 118109, 117008 th rough ' 117010, 117013, 120782,
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.1217?1 Vendor Manuals:
M2.02-45, M19.02-313, M21.08A', N6.03 IMO, 30120-0M-01 & -02 (Tec Accelerometers).
B&W Site Bulletin 81-076 Dresser Report SV214, Rev. 0 E&QC Memo 81-38 9
ASME III, Article 9, 1968 Nameplate Data Spare Parts Listings Equipment History IE Notice 86-05 DG Memo 86-272 Design Specifications Design Calculations:
1 2-IAS-M2084, Rev. 2, !
Z-IAS-M2085, Rev. 2, 2-MSS-M0433, Rev. 3, I-MSS-M0431, Rev. 4, Z-MSS-M2055, Rev. O, Z-MSS-M0829, Rev. O ATTACHMENT 1
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~# LIST OF REVIEWED DOCUMENTS (Continued)
Davis Besse TAP Report on- March 2,' .1984 ' Trip -
Regulatory Guide-f 10 CFR 50, Appendix A, B & R NRC'ASRTP Report Memo GVC-87-681 I&C Calibration'& Maintenance Records Precursor Review. Task Final. Report Deterministic Failure Consequences Recommendation 0588 i
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ATTACHMENT 1 i
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" STATUS OF RIs h
STATUS RSL NUMBER RI NUMB.ER 010 Open RSL-RI 010 014 Open RSL-RI 014 01 5 Open RSL-RI 015 019 Acknowledged RSL-RI 019 037 Open RSL-RI 037 ,
046 Acknowledged RSL-RI 046 047 Open RSL-RI 047 066 Open RSL-RI 066 097 Open RSL-RI 097
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ATTACHMENT 2 s
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- 2. ,
, p j. EXPANDED AUGMENTED SYSTEM REVIEW AND TEST PROGRAM (EXPANDED ASRTP)
EVALUATION ,
OF THE
- INTEGRATED CONTROL-SYSTEM SUBMITTED BY: < c <f / . v- DATE: 68-C/-67 RICHARD M0YER /
TEAM LEADER L
CONCURRENCE: l t b% DATE: ' f'O '
V10 HUMENANSKY-EXPANDED ASR PROGRAM M ER CONCURRENCE:
/ DATE:
/
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/ BOB CROLEY / \
DIRECTOR, NUCLEAR TECflNICAL SERVICES
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-c, TABLE 0F' CONTENTS .
Page Number
- 1. 0 ' ' INTRODUCTION. 3 l
2.0 PURPOSE 4 c .
l 3.0 SCOPE ." 5 4.0, OVERALL RESULTS AND CONCLUSIONS. 6-8
- 5.00ETAILED OBSERVATIONS - REQUESTS'FOR INFORMATION 6.0 ATTACHMENTS- 23 -
l 6.1 List.of Documents Reviewed
.' 6. 2 ' Status of RIs q
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e-EXPANDED AUGMENTED SYSTEM REVIEW AND TEST PROGRAM
..y EVALUATION OF THE INTEGRATED CONTROL SYSTEM l
- 1. 0' INTRODUCTION The Rancho Seco Expanded Augmented System Review and Test Program
[ASRTP] evaluation ef fort involves an assessment 'of the effectiveness of the System Review and Test Program [SRTP] and an analysis of' the adequacy of 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 implemented on 33 selected systems and an.in-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 mul;i-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 g members with consultants, architect engineer and vendor backgrounds-are joined with the specific plant knowledge of SMUD team members.
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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 enhance the evaluation. Each team prepares a final report for each completed selected system evaluated. This report is for the Integrated Control System (ICS).
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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 effe-tiveness of established programs for ensuring
, safety during plant operation af ter restart.
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/ 3.0 SCOPE To accomplish the first objective, the Control Systems team.
evaluated the Integrated Control System to determine whether:
- 1. The system was capable of performing the functions required by its design bases. '
2 Testing was adequate to demonstrate that the system would perform all of the functions required.
- 3. System maintenanceiwas adequate to ensure system operability:under postulated accident conditions.
- 4. Operator and maintenance technician training was adequate to ensure proper operations and maintenance of I 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 Control Systems team reviewed the programs as implemented for the Integrated Control System in the following functional areas:
- 1. Systems Design and Change Control
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- 2. Maintenance
- 3. Operations and Training
- 4. Surveillance and Inservice Testing
- 5. Quality Assurance
- 6. Engineering Programs The Control Systems team reviewed a number of documents in preparation for and during the Expanded ASRTP evaluation. This list of documents is found in Attachment 1.
The primary source of leads for the team were the problems identified in the ICS 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 Integrated Control System included a review of pertinent portions of support systems that must be functional in order for the Integrated Control System to meet its design objectives.
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.: 14 . 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 ' af ter . restart are summarized below. The summary focuses on the weaknesses identified during the evaluation. Section 5.0 provides detailed findings by providing.the Request for Information (RI) forms that are used by the Expanded ASRTP teams to identify J potential = concerns during the evaluation. The numbers in brackets after each individual summary refer to the corresponding RIs in ,
Section 5.0.
4.1 . Engineering Design 4.1.1 ICS devices may be calibrated to incorrect setpoints. ICS !
devices are presently. set using information from calibration I records which are not controlled documents. The calibration -
records provide no source for their setpoint content and no mechanism exists in design modification procedures to ensure setpoint changes are documented and incorporated into the plant design. Incorrect-setpoint actuations could result in , !
unexpected plant control or prevent actuations from 1 occurring at their required design values. (RI-036) 4.1.2 The intended sequential operation of Turbine Bypass Valves (TBVs), Atmospheric Dump Valves (ADVs) and Main Steam Relief q Valves (MSRVs) may not be achievable due to possible overlapping of the actuation setpoints as a result of setpoint tolerances. Of particular concern is the possible opening of ADVs prior to the TBVs in the event of tube rupture which could cause unnecessary atmospheric releases.
Compounding the problem is a large number of discrepancies 1 in the setpoints and their tolerances due to a lack of a single setpoint source document (Ref: Item 4.1.1) and calculations which lack consistency in approach and cross referencing of source information. (RI-028) ;
4.1.3 Inadequate design analysis and documentation exists for '
determination of adequate sizing of Integrated Control System (ICS) breakers lJ04 and 1C07. ICS System Status ,
Report Problem No. 18 discussed discrepancies in the two j breaker sizes (30 am] vs 40 amp) and provided a resolution of analysis to determine the correct breaker capacity.
However, the conclusion of the resulting analysis did not 1 support the replacement of the 30 amp breaker with a 40 amp j breaker and did not determine actual loads in reaching the ;
conclusion. This may reduce plant availability / reliability I and may damage equipment if the breaker is oversized.
(RI-096) 6 I
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- OVERALL RESULTS AND CONCLUSIONS (Continued) l 4.2 Programmatic Cor.cerns 1 1
4.2.1 ' Engineering : Change-Notices (ECNs) are being closed without all modifications being performed, which could lead operators to take, inappropriate or ineffective actions. ICS System Status Report Problem #46 identified the necessity of
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replacing shunt trip switch labels for ICS and NNI. The ECN incorporating the modification was closed without the labels for NNI being replaced. (RI-039)-
4.2.2 No program exists to recall or void active work reque'sts for Preventive Maintenance (PM) tasks when a new or modified PM task is entered into the system. PM tasks are not identified by revision, and the potential exists for the earlier work request based on the superseded PM task to be worked after the revised PM task work request which could cause incorrect maintenance to.be performed. (RI-074) 4.3 Testing 4.3.1 The ICS functional test appears inadequate to verify errors in the system which may have resulted due to.the massive ~
rework of terminations resulting from an identified cause of ;
the December 26, 1985 trip as noted in ICS System Status Report Problem #10. Quality Control inspection was requested and performed on the quhlity of the terminations only. Recent documented findings show that some terminations were not made to their correct locations. This is of particular concern since the team noted that a plant trip after restart resulted at 0 avis-Besse f rom an ICS wiring discrepancy which was undisclosed during ICS functional testing. (RI-080) 1 7
i 5.0' 0ETAILEP OBSERVATIONS - REQUEST'FOR INFORMATION
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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 ;
M the EASRTP team is having difficulty obtaining.
These RIs are considered drafts throughout the entire evaluation until they become part of the final 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 report.
Attachment 2 of the report provides RI status as of this report-d Ce. 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. l Approximately one week will be provided after the report is~ issued to provide time for departments to address each RI for va.lidity. A :j revision to Attachment 2 will then be issued to reflect the status of RIs. All RIs not acknowledged at the end of this period will have an "Open" status. RIs are then transferred into the Restart Scope List tracking system for resolution and. corrective action i' implementation.
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L REQUEST FOR INFORMATION'(RI)
RI NO: 028 SYSTEM CODE: ICS ISSUE DATE: 7-23-87 l
SUBJECT:
-SETPOINTS FOR THE STEAM BYPASS SYSTEM DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR: RON LAWRENCE TEAM MEMBER: SINGH BAGGA TEAM LEADER: RICH M0YER POTENTIAL CONCERN /00ESTION:
The intended steam bypass system operation according to the Main Steam System (MSS) design bases document is to open Turbine-Bypass Valves (TBVs), Atmospheric Dump Valves (ADVs) and Main Steam Relief Valve (MSRV) sequentially in this order which may not be achievable due to the overlap of set points (S.P.) for these valves when the set point tolerances are taken into consideration. The overlapping of the set points could cause !
ADVs to open prior to opening T8Vs which could release unnecessary.
radioactivity to the atmosphere in the event of tube rupture.
The plant personnel were unable to provide the documents supporting the selection of the setpoints for-the TBVs and ADVs,for review. Several discrepancies were noted among the documents for the setpoint values and the setpoints are not controlled through any established method or g procedure within the organization.
- OVERLAP OF THE SETPOINTS FOR TBVs. ADVs AND MSRVs:
- TBVs S.P. (psig)
- ADV's S.P. (psig) * .First MSRV S.P. (psig) *
- Nominal
- Min
- Nominal.* Max
- Min **Nominal * *Max *
- 1000
- 990-* 1020 *1050
- 1035
- 1050
- 1050
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- ICS SDB P.13
- Calculation #
- Process standards
- AP-150 p.16
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- Table 3
- Z-EFI-IO146 p. 6
- DISCREPANCIES IN ADVs SFPTOINT LOOP ERRORS:
After reviewing the calculations Z-EFI-10146 Rev. O and Z-EFI-10157 Rev. O, Process Standards AP-150 Rev.14 and System Design Bases (SDB) for ICS Rev. 1 draft, the following concern is noted:
Calc Z-EFI-IO146 Pages 6 & 43 1020.117 psigt30.05 psig i Calc Z-EFI-10157 Page 2 1020 psigi25 psig l 9
.- POTENTIAL CONCERN /00ESTION:
Af ter. reviewing ICS SSR Rev.1, System design basis for ICS Rev.1 draf t and MSS Rev.1 draf t, System Description N21.01-112 Vol. 2 and Plant Operating Procedures B.2 Rev. 38, Engineering Change Notice (ECN) A-5415 Rev.1, the following discrepancies were noted:
. DISCREPANCIES IN SETPOINT FOR THE TBVs CONTROL ON REACTOR TRIP:
ICS SSR p. 2-7 para 2.2.2.d 1015 psig i
ICS SDB p.13 Table 3 1000 psig MSS SDB p. 5 para 2.2 125 psi above normal S.P. of 885 psig=1010 psig 8 & W ICS final documentation 125 psi above normal s.p.
Sheet IC 5 of-8 System description p. 3.4 115 ps; above normal S.P. of 885 psig=1000 psig . ,
ICS Schematic Diagram N21.01-72 Rev. 6 115 psi above normal S.P.
ICS Calibration record (
pp.129 & 130 115 psi above normal S.P.
. DISCREPANCIESINSETPOINTFORTHETBVsC0kTROLONTURBINETRIP:
MSS SDB p. 5 para 2.2 pressure control point i biased by small amount ICS SDB p.13 Table 3 setpoint changes from 935 psig to 885 psig
. DISCREPANCIES IN SETPOINT FOR THE TBVs CONTROL DURING NORMAL ;
i OPERATION:
-l ICS SDB p. 13 Table 3 935 psig MSS SDB p. 5 para 2.2 885 psig l
Plant Operating Procedure B.2 l Rev. 38 p. 29 935 psig (885+50) )
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I. : POTENTIAL CONCERN /00ESTION:
. ADV CONTROL:
MSS SDB pp. 30 & 31 para 6.2.1 ICS Controls ADVs
-105 508 p.13 Table 3 ICS Controls ADVs ECN A-5415 Rev.1 Emergency Feedwater Initiation and Controls (EFIC) Controls ADVs l
- . TBVs OVERRIDE:
ICS SDB p.13 Table 3 1035 psig Calibration Record: 1050 psig Af ter reviewing calculations Z-EFI-10146 Rev. O and 2-EFI-10157 Rev. O, the following discrepancies were noted:
. 7-EFI-10146 p. 6 item 3.a . ADV Setpoint is 1020.117" pp. 35 & 43 . Max. error in ADV control loop 130.45 psig vs 130.05 psig.
- p. ,43 f.1.j . EADVL-1200 x 2.5%= 30.05 psig
- p. 36 & 39' . Error direction (+ or -) not called out consistently.
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. 7-EFI-10157 p. 22 . Assumption for the error is 2%
of Upper Range Limit (URL) but calculation uses 2% of-calibrated span.
. Assumption for the errors in Buffer and Control Module is not considered.
. Reference for the nominal value of setpoint is not called out.
. Error direction (+ or -) not called out consistently. l I
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. ADV CONTROL LOOP ERROR INCONSISTENT:
l Z-EFI-10146 pp. 35 & 43 - Total loop error 130.05 psig i Z-EFI-10157 pp. 2 & 22 Total loop error 125 psig
. The calculations are not. cross- l 1
referenced (in Z-EFI-10157) 11 P
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e, REQUEST FOR INFORMATION.(RI)
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~ RI NO: 036 SYSTcM CODE:- ICS ISSUE DATE: 7-27-87 i
SUBJECT:
CONTROL SETPOINTS- DOCUMENTATION NUCLEAR ENGINEERING COORDINATOR: RON LAWRENCE DEPARTMENT:
JEFF IRWIN TEAM LEADER: RICH M0YER
- TEAM' MEMBER:
POTENTIAL' CONCERN /0VESTION:
Integrated Control System devices may be calibrated to incorrect-
, setpoints.
- 1. ICS'setpoints are not in process standards and the plant could not produce.another controlled setpoint document. I&C shop calibration records appear to have complete electronic hardware settings and are archived but not controlled.
- 2. Programmatic control of implementing new or modified setpoints is unclear. Review of procedures for-generation, implementation and
. turnover of. modification packages (NEP-4109, RSAP-803 'and AP44) revealed no directions for addressing setpoints. There is'no guidance on analyzing,. documenting and transmitting design setpoints-and tolerances. There is: no flow chart for: implementing design
.setpoints and disseminating setpoints to setpoint documents and training . . \ .,
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- 3. '. Plant was unable to provide cross reference to, or assurance ofexi 1 providing uncertainty in meeting USAR, Technical Specifications, Regulatory and functional' requirements for ICS and other plant systems.
- 4. No controlled documentation of module functions exists. Maintenance and modifications require detailed knowledge of signal functions l
including engineering units.
S. Process Standards is a partial setpoint document of selected ~
instruments. It does not reference support documents and does not provide signal to engineering unit conversions.
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.- REQUEST FOR INFORMATION (RI)
RI NO: 039 SYSTEM CODE: ICS ISSUE DATE: 7-27-87
SUBJECT:
LABELING OF NNI SHUNT TRIP SWITCHES-ECN CLOSURE .
DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR: RON LAWRENCE i TEAM MEMBER: P. WAGNER TEAM LEADER: RICH M0YER POTENTIAL CONCERN /0VESTION:
Engineering Change Notices (ECNs) are being closed without all modifications being performed, which could lead operators to take inappropriate or inef f ective actions.
. According to the ICS System Status Report Problem #46, S1 and 52 Shunt Trip Switches were not labeled to provide positive indication ,
that the switches were in a tripped condition.
. According to the ICS System Status Report, "ECN R-0472 relabeled breakers S1 and 52 on the ICS and NNI so that positive indication of these breakers being open can be determined without question."
. ECN R-0472 Rev. O specified relabeling S1 and 52 on the ICS alone. g (ECN is closed)
. ECN R-0580 Rev. O specified relabeling sl\and 52 on the NNI. The Design Basis Report states: "In addition labels will be added to S1 and S2 in the X and Y power distributions to aid the operator in determining switch position." (ECN is closed)
. The plant could not produce another ECN that labels NNIs S1 and 52.
. A walkdown of the NNI verified that labels have not been added to NNIs Si and 52.
. Work Request 114092 included replacing NNI S1 and 52 labels but was closed without the work being performed. 4
. There is a potential that the operator, attempting to find the cause of an NNI loss of power may not be able to identify that the S1 and S2 switches are tripped. This happened on identical switches in the ICS on December 26, 1985.
. ECN R-0580 was released to operations without all work specified in the ECN completed. This does not meet the intent of AP44 Plant Modification procedure which states prior to ECN release installation, testing, and design of the modification must be complete. Contributing to this concern is that a work request had been written to cover the work in question and was signed as completed, but the total work was not performed.
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REQUEST FOR INFORMATION (RI)
RI NO: 043 SYSTEM CODE: ICS ISSUE DATE: 7-24-87
SUBJECT:
ICS SHUNT TRIP SWITCHES S1 AND S2 PERFORMANCE LIFE EXPIRES PRIOR TO REPLACEMENT .
DEPARTMENT: MAINTENANCE COORDINATOR: JIM DARKE _
TEAM MEMBER: P. WAGNER TEAM LEADER: RICH M0YER POTENTIAL CONCERN /0UESTION:
ICS and NNI shunt trip switches (S1 and S2).may not be functional, prior to' scheduled replacement because the Preventative Maintenance (PM) Task does not account for shelf life.
. -The manufacturer of the= switches stated that they would not
. guarantee the switches for longer than five years after the switch leaves the factory. (per Engineering Report (ERPT I-0010)
. The PM Task (05608) for replacement of the switches has a performance interval of 1825 days (five year,s) and does not account for shelf life. .
5 Therefore, since shelf' life is not considered, a period will exist when the switch is installed'in the field and is no longer guaranteed by the manufacturer. Failure of these switches could,mean a spurious loss of ICS power or a failure to trip ICS power.
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. REQUEST.FOR INFORMATION (RI) i.
RI NO: 073 ' SYSTEM CODE:- ICS ISSUE DATE: 7-29-87
SUBJECT:
USAR DESCRIPTION OF :ICS 3-DEPARTMENT: LICENSING _
COORDINATOR: J. DELEZINSKI TEAM MEMBER: JEFF IRWIN TEAM LEADER: RICH MOYER POTEN'TIAliCONCERN/0UESTION:
USAR~ description of functioning of ICS control stations on loss of power ,
.is incorrect.
. USAR, including amendment 5 revisions, section 7.2.3.3.1, is incorrect in stating " Loss. of electrical power to automatic control stations reverts the control system to Manual."'
. .The Bailey 720 design such as Oconee and TMI operate in this manner, but the Bailey 820 design used at Rancho Seco has both automatic and manual control being powered from the same DC source.
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t *f REQUEST FOR INF0PMATION (RI)
RI NO: _0_74 SYSTEM CODE: ICS - ISSUE DATE: 7-29-87
SUBJECT:
REVISIONS TO PM TASKS _
DEPARTMENTS MAINTENANCE COORDINATOR: J. DARKE' TEAM MEMBER: JEFF IRWIN TEAM LEADER: RICH M0YER _
POTENTIAL CONCERN /0VESTION:
-Superseded Preventive Maintenance (PM) tasks may be worked even though_
the respective new or modified PM tasks have been entered into the system.
.. No procedure's (ref. AP650) exist'to recall active work requests written by PM tasks when a new or modified PM task for the respectite device is entered into the PM system.
. PM tasks are not identified by revision number.
+ As . rcsult wo"k requests incorporating old directions may get worked after the revised modifications work request, causing incorrect maintenance to be performed.
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2 j REQUEST FOR'INFORMATION (RI)
L RI NO: 080 SYSTEM CODE: ICS ISSUE DATE: 7-29-87
SUBJECT:
TESTING OF WIREWRAP REWORK IN THE ICS DEPARTMENT: SYSTFM ENGINEERING COORDINATOR: JOHN ITTNER TEAM MEMBER: PAT WAGNER TEAM LEADER: RICH M0YER POTENTIAL CONCERN /0VESTION:
The ICS functional test appears inadequate to detect errors resulting from termination rework; errors which could lead to a plant trip or to making a plant trip more complex.
. Due to an identified cause of the December 26, 1985 trip, a bad lug in the ICS, a significant number (thousands) of terminations and wirewrap connections in the ICS were inspected and reworked. (ICS System Status Report Problem 10)
. Quality Control inspection was requested and performed on the quality of the terminations only. Recent findings (Nonconformance Reports S-6603, S-6867) indicate some tern.inations were not returned
.to their proper-location, g
. A-discussion with an I&C Maintenance Engineer indicated that maintenance is relying on the ICS Functidpal Test (Special Test Procedure STP.778) to find any discrepancies.
. Although the ICS Function Test (STP.778) tests the functions of the ICS, the test does not verify that every wire in the ICS is landed at the-proper terminal.
. A review of a section of the ICS Function Test (Section 6.4, Feedwater) showed that only approximately 80% of the wiring in the subsystem will be tested.
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. The B&W ICS Engineer who was involved in the Davis-Besse restart stated that a plant trip af ter-restart at Davis-Besse was attributed directly to a wiring discrepancy which was missed during function testing.
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. REQUEST'FOR INFORMATION (RI)
RI NO: 081 SYSTEM CODE: .ICS ISSUE DATE: 7-30-87
SUBJECT:
ICS/NNI ABT OPERATION DEPARTMENT: OPERATIONS COORDINATOR: R. MACIAS TEAM MEMBER: JEFF IRWIN TEAM LEADER: RICH M0YER POTENTIAL CONCERN /0VESTION: .i
. The Automatic Bus Transfer ( ABT) is designed such t' hat a situation could exist where ICS AC loads are lost but ICS DC loads remain energized causing uncertainty in ICS control.
. Review of drawing E203 Sht 99 Rev. 2 shuWed ICS ABT performs automatic bus transfer f rom primary AC power alignment to alternate AC power alignment; but will n_o_t perfcrm automatic bus transfer f rom alterncte to primary.
. The operator must manually pu:sh the reset button located on the ABT enclosure, after primary AC power is rrstored, to realign the ABT to primary AC power.
. Plant could not provide a design basis requiring use of a unidirectional versus a bidirectional power seeking ABT.
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. ABT alignment to alternate AC power is annunciated. . Annunciator Response Procedure (ARP) H2PSB addresses the need to reset the ABT to primary AC power alignment. Plant conditions may not allow immediate/near term realignment to primary AC power.
. Plausibly the ABT could be aligned to alternate AC power, the ABT primary AC source would be restored and prior to the time the operators reset the ABT to primary AC power alignment the alternate AC power could be lost. This would de-energize all ICS AC loads while all ICS DC loads would still be energized.
. Operators interviewed knew that the ABT was unidirectional power seeking. Operators were not aware that there was a possibility that ICS AC loads could be lost while ICS DC loads remained energized.
This condition would cause loss of MFW pumps while ICS controllers remained energized and ICS control actions would become unpredictable.
. The plant could not provide an analysis of ICS with AC loads lost and DC loads energized.
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REQUEST FOR INFORMATION (RI)
RI NO: 093 SYSTEM CODE: ICS ISSUE. DATE: 7-30-87
SUBJECT:
LABELING 0F AUXILIARY STEAM STATION FOR LOSS OF:ICS POWER DEPARTMENT: SYSTEM ENGINEERING COORDINATOR: JOHN ITTNER TEAM MEMBER: PAT WAGNER TEAM LEADER: RICH MOYER POTENTIAL CONCERN /0VESTION:
On a loss of ICS power, operators may not be aware that Control Room control of Auxiliary Steam is not functional.
ICS System Status Report (SSR) Problem 22 identified that all Control Room instrumentation and controls which are powered from ICS should be labeled. A portion of the Auxiliary Steam Reducing Station in the Control Room is powered from ICS. There is no label in the Control Room indicating that the station is powered from ICS.
. During the ICS Equipment Investigation ( Action List Item Number 3A),
upon loss of.ICJ power, it was noted that the Auxiliary Steam Reducing Station lost power unexpectedly.
. In the resolution of ICS SSR Problem 22, ICS Bailey Hand / Auto (H/A)
Stations did not require labeling. The ICS Bailey Hand / Auto stations are grouped together on panel Hlhi and, being ICS stations, do not require labeling indicating that they are ICS powered.
. The Auxiliary. Steam Station, at one time a Bailey station, is not located among the rest of the H/A stations and is no longer clearly identified as being powered by ICS (ECN R-0878).
. A walkdown of the Control Room confirmed that there is no purple l label signifying that this controller is af fected by a loss of ICS power.
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,f REQUEST FOR INFORMATION (RI)
RI NO: 094 SYSTEM CODE: ICS ISSUE DATE: 7-30-87
SUBJECT:
EFFECTIVENESS OF ENGINEERING REGARDING ECN WOR 4 PACKAGES I
COORDINATOR: RON LAWRENCE DEPARTMENT: NUCLEAR ENGINEERING TEAM MEMBER: JEFF IRWIN TEAM LEADER: RICH MOYER POTENTIAL CONCERN /0VESTION:
Engineering design walkdowns and verifications for design modifications as found in NEP 4109 are not being performed adequately.
Review of Engineering Change Notice (ECN)-work packages (R-0861 and R-0927) in progress showed a large number (25 and 26 respectively) of Field Problem Reports (FPRs) being generated. The majority of FPRs appeared to be of technical nature requiring design correction. There is.
a concern that design errors that affect functionality of the system may avoid detection by installation personnel and modification testing.-
Review of FPRs for ECN R-0861 revealed several instances where the original design called for instrument cables to be routed in trays that the field determined was unacceptable. Field engineering then ;
recommended the cables be routed in trays which were power trays, i Engineering corrected this error in the FPR before responding, but this scenario reflects the necessity for engineering walkdowns to be performed initially per the procedures . Engineering's initial response to this concern stated that drawing details.were difficult to read or lacking in detail, which also emphasizes the need to perform walkdowns and verifications before completing the design.
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w REQUEST FOR INFORMATION (RI) k RI N0: 095 SYSTEM CODE:- ICS ISSUE DATE: 7-30-87
$UBJECT: CALIBRATION RECORDS CONTROL i
DEPARTMENT: MAINTENANCE COORDINATOR: J. DARKE , e
. TEAM MEMBER: JEFF IRWIN TEAM LEADER: RICH M0YER I
POTENTIAL CONCERN /00ESTION: ,
No procedures exist for I&C maintenance to ensure calibration data sheets ~
are microfilmed in accordance with Records Management requirements. j
. In Response to NRC open item 86-22 calibration data sheets were brought into compliance of permanent plant records by microfilming ]
in early 1987.
. Interview with I&C shop supervisor indicated that new data sheets- -j are collected.and periodically microfilmed by records management to l maintain permanent plant record status, by arrangement with Records management.
. Review of Maintenance Procedures 1011 Rev.3 and 1014 Rev. 2'd k not l' show programmatic control of this process of micro filming all new j calibration data. No occurrences of calibration data sheets not being microfilmed or calibration data shhets being lost were found.
However, there is a potential for these occurrences with no procedural control of this activity at the shop level under the i I
'RSAP-0601 guidelines.
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.. REQUEST FOR INFORMATION (RI)
RI NO: 096 SYSTEM CODE: ICS ISSUE DATE: 7-30-87
SUBJECT:
ICS BREAKERS FOR 118VAC POWER SUPPLIES DEPARTMENT: NUCLEAR ENGINEERING COORDINATOR: RON LAWRENCE TEAM MEMBER: SINGH BAGGA TEAM LEADER: RICH M0YER POTENTIAL CONCERN /0VESTION:
The ICS breakers 1J04 and 1C07 may not be sized properly. This could cause unnecessary /f requent tripping of the breakers or could cause damage to the equipment if the breakers are oversized, thereby reducing plant availability and reliability.
. The plant is unable to provide a document. showing all the loads to support the sizing and selection of breakers. The loads may be shown on several drawings but it is very difficult to verify the loads due to several modifications being made to ICS.
. The engineering report ERPT-E0208 Rev. O still shows various loads which have been removed from ICS. It appears from this report that the breaker loads do not get re-analyzed or updated. g
. Drawings E-1011 Sh.110-2 and Sh.115A inqicata the " total load" of 25 Amps for the breakers but the individua'l loads are not called out on this drawing. Discus's ions with engineers revealed that the
" total load" of 23.15 Amps is identified in the instruction manual for ICS. However, the instruction manual never gets updated for the revised loads. In addition, it was stated that it is good engineering practice to select a breaker of 125% capacity of ful' load. The 30 Amp breaker 1J04 has been replaced with a 40 Amp breaker under Engineering Change Notice (ECN) R-0469 Rev. O.
i
. The calculation #Z-ICS-E0640 Rev. 1 was done for the coordination of breakers and fuses of 118VAC circuits. It was concluded from this calculation that fuses in 118 VAC circuits coordinate properly with 40 Amp and 30 Amp breakers for all f ault currents available f rom the inverter or f rom the regulating transformer. This conclusion did not support the replacement of 30 Amps breaker under ECN #R-0469 Rev. O.
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.e -6.0 ATTACHMENTS 6.1 List of Documents Reviewed e.2 stetus e, m 1;
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42 LIST Of DOCUMENTS REVIEWED
' ATTACHMENT I ;
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1.: 'SSFIs A. ,ANO B.. Oconee 4 C. Palisades D. Pilgrim E. Rancho Seco. .
F. H.R. Robinson G. Three Mile Island 1 H. Trojan
.I. Turkey Point
'23 System Status Reports
.A. Integrated Control System (revision 1)
B. . Main Steam System (revision 1)
C. Main Feedwater (revision 1) ,
D. Non-Nuclear Instrumentation (revision 1)
-3. SMUD Vendor Technical Manuals
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A .' N21.01-115 (ICS Volume 1 A)
B. N21.01-116 (ICS Volume 18)
C. N21.01-117 (ICS Volume 1C)
D. N21.01-112 (ICS Volume 2)
E. E32.09-IM01 ( Automatic Buss Transfer Devices)
- 4. Drawings A. N21.01 series (ICS Schematics)
B. E203 shts 99(r2),100,101 C. M-530 shts 1 (r13), 2 (r12), 2A (r3), 3 (r17)
D. M-532 shts 2 (r6), 3 (r6)
E. Pink Drawings in N21.01 series for ECNs:
(1) A-4058 (2) A-5415W,Y,2 (3) R-0823 (4) R-0824 !
(5) R-0825 (6) R-0826 l (7) R-0861 i (8) R-0878 (9) R-0927 (10 R-1217 24
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. I I ,- LIST OF REVIEWED, DOCUMENTS ATTACHMENT 1 (Continued)
- 5. Updated Safety Analysis Report Chapter 7 (ICS) .
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- 6. Nonconforming Reports A. S-5263 B. 5-6603 C. S-6867 1
- 7. Licensee Event Report ;
A. 86-10 B. 85 8. Quality Control Instrumentation 12 (QCI-12), r2 i
- 9. -Nuclear Engineering Procedures 1
A. NEP-5435 ICS System Design Bases, r1 B. NEP-5100 Series Construction Specs ;
C. NEP-4109 Configuration Control, r6 }
- 0. NEP-6118 E. NEP-5444 MSS System Design Bases, r0
- 10. ICS. Testing Documents
'A. System. Test Matrix B. Test Outlines for:
(2) STP 664 r0 (Loss of ICS, NNI)
(3) . STP.778 r0 (ICS Function Test)
C. Test Procedures for:
(1) STP.660 r0 (2) STP.778 r0 3 4
- 11. Procedures A. Administrative Procedures {
(1) AP.4A, Safe Clearance Procedure, r5 I
l (2) AP.44, Plant Modification, r11 (3) AP.100 series (Process Standards), r11 l
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g g- LIST OF REVIEWED DOCUMENTS ATTACHMENT 1 .(Continued)
B .- Rancho Seco Administrate,ve Procedures (1) RS AP. 601'.
(2) RSAP.803 C. Operating Procedures (1); A.71 kCS Operating Procedure, r9
'(2). B.2, r38 (3) 8.3, r29
.(4) B.4, r40 D. Casualty Procedures (1). C.13A,.r6
'(2) C.138, r4 (3) .C.40 Loss: of- ICS Power, r2 draf t
'E. Annunciator Response Procedure
-(1)~ ARP.H2PSB, r15 g F. Modification Procedures and Inspection Standards 302
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- 12. Engineering Changes'(ECNs, DBRs). ;
l A. R-0357A,B B. R-0359A,-B j
C. R-0459- !
D. .R-0472 E. R-0477 j
F. R-0580 i
G.
H.
R-0822 R-0823 j
R-0826 l I.
j J. R-0828 K. R-0861 L. R-0878 M. R-0918 N. R-0927 O. R-0442 P. R-0469 :
- 13. Field Problem Reports- ;
A. R-0927 FPR1-FPR26 B. R-0861 FPR1-FPR25 26 1
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.. LIST OF REVIEWED DOCUMENTS ATTACHMENT 1 ' ontinued)
.14 PM Tasks for the ICS l A. Task 04145 B. Task 05355 C. Task 05378
- 0. Task 05389 i E. Task 05602 F. Task 05608 G. Task 05745 H. Task 05746
- 1. Task 05749 J. Task 05750
- 16. Work Request History for ICS
- 18. Standard ANSI Nk45.2.11 1974
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- 20. Engineering Reports A. ERPT-10010 (SAIC Report)
B. ERPT-E0208 (Fuses for 118 VAC) >
- 22. CRTS letter GCA 87-090 dated 7/21/87
- 23. BAW-1919 Draf t, BWOG I&C Committee, ICS/NNI Evaluation, Appendix R
- 24. Engineering Calculations A. 2-ICS-E0640 B. Z-ICS-E0598
! C. Z-EFI-10146 D. Z-EFI-10159 E. Z-EFI-IO158 F. Z-EFI-10163 27 B
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- LIST OF REVIEWED DOCUMENTS ATTACHMENT 1 (Continued) l 1
' 25. Training Materials 1
A. Nuclear Training Department Course' Catalog (TTM07, 7/16/87)
B. Lesson Plan, ICS (00 21 I 4900 rl)
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' .4.
8 STATUS OF RIs r.
RI NUMBER STATUS RSL NUMBER 028 Acknowledged RSL-RI-028 036 Acknowledged RSL-RI-036 039 Acknowledged RSL-RI-039 043 ,
Acknowledged RSL-RI-043 073 Acknowledged RSL-RI-073 074 Acknowledged RSL-RI-074 080 Open 081 Acknowledged RSL-RI-081 093 Open 094 Open 095 Open 096 Open
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ATTACHMENT 2 1
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