05000416/LER-1986-029, :on 860826,discovered Incorrect Flow Measuring Device Coefficient Used in Establishing Coolant Flow to ESF Electrical Switchgear Room Coolers Resulting in Low Flows. Caused by Ae Design Error.Pipes Flushed
| ML20237H495 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 08/28/1987 |
| From: | Byrd R, Kingsley O SYSTEM ENERGY RESOURCES, INC. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| AECM-87-0169, AECM-87-169, LER-86-029, LER-86-29, NUDOCS 8709030297 | |
| Download: ML20237H495 (26) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability |
| 4161986029R00 - NRC Website | |
text
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Pipe Fouling Results in Low Flows to ESF Room Coolers and Related SSW Design Concerns SV8WT DAff (S)
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REPORT DATE (7)
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On August 26,198f, it was discovered that an incorrect instrument flow coefficient was used in establishing coolant flow to the Engineered Safety Feature (ESF) electrical switchgear room coolers in the Fall of 1985. During investigation of that incident, system checks revealed that with the system in the post-LOCA configuration, Standby Service Water (SSW) flow rates to some of the Division 1 and Division 2 ESF switchgear room coolers were below design I
l values due to the partial blockage of small diameter branch piping supplying l
the cooling water to the cooling units. Additional investigation into the concern revealed that the flows to the "A" control room A/C unit and RCIC room cooler were below the original established flows, that one ESF switchgear room cooler was apparently undersized, and that the allowable nozzle loads for the, coil connections on 18 safety-related room coolers had not been appropriately considered in the original design. The investigation also revealed inadequate engineering assumptions regarding atxiliary building corridor post-accident temperatures.
The low flow concerns were resolved by flushing or hydrolazing pipes, system modifications, increasing fluid flow velocities, improving chemical treatments, and establishing a long term flow monitoring program.
Piping support modifications were made to resolve the nozzle load concerns. Using analysis the affects of the increase corridor temperatures were determined to require no hardware changes.
A special evaluation of the SSW system has revealed that further design h
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deficiencies exist. These are discussed in Section G of this report.
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REPORTABLE OCCURRENCE 1.
Inadequate Cooler Flow On August 26, 1986, it was discovered that an incorrect instrument flow coefficient was used in establishing coolant flow to some of the Engineered Safety Feature (ESF) electrical switchgear room coolers in the Fall of 1985.
During the investigation of that incident, system checks revealed that with the system in the post-Loss of Coolant Accident (LOCA) configuration Standby Service Water (SSW) flow rates to some of the Division 1 and Division 2 cooling units were below post accident design values.
The flow impediment was due to the partial blockage in sections of small diameter branch piping supplying the cooling water to the cooling units. Additional investigation into this concern revealed that flows to the "A" control room A/C unit and RCIC room cooler were also below the original established flows.
In the coun e of re-evaluating room heat loads, the engineering staff also determined that the Division 2 ESF switchgear room cooler at El. 119' East had insufficient heat removal capability.
2.
Inadequate Cooler Nozzle Support While performing pipe support analyses in conjunction with corrective actions for the cooler mentioned above (El 119' East), the engineering staff determined that the allowable nozzle loads for the coil connections on 18 safety-related room coolers had not been appropriately considered in the original design thereby causing their operability during a postulated seismic event to be questionable.
The above items represent a condition which is reported pursuant to 50.73(a)(2)(v)and50.73(a)(2)(i)(B).
A special evaluation of the SSW system has revealed additional reportable occurrences which are discussed in Section G of.this report.
These additional occurrences are reported pursuant to 10CFR 50.73(a)(2)(v) and 10CFR50.73(a)(2)(vii).
B.
INITIAL CONDITIONS The plant was in Hot Shutdown at the time of the discovery of the initial reportable condition. Additional findings were identified during the first refueling outage.
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DESCRIPTION OF OCCURRENCE 1.
Inadequate Cooler Flow On August 26, 1986, during the development of special instructions to test and flow balance the SSW "A" Loop following system modifications scheduled for GGNS first refueling outage, it was discovered that an incorrect flow measuring device coefficient was used to flow balance certain components in the B loop in November 1985 following modifications to the B loop required by Operating License L
Condition 2.C(20). Additional investigation revealed that the loop l
flow to the "A" ESF switchgear room coolers was also verified in December 1985 using the same incorrect #10w coefficient. The type of flow measuring device used in the flow balance was an Annubar Type 73.
A flow coefficient of 0.9779 was used for the device rather than the correct coefficient of 0.5710.
A system flow check was conducted with the system aligned in the post-LOCA configuration. The checks revealed that the SSW flow rates to the Division 1 and Division 2 ESF switchgear room coolers were below post accident design values due to flow impediments caused by the fouling of small diameter branch piping supplying the cooling water flow to the cooling units. The cooling units are designed to maintain the ESF switchgear room temperature less than 104 degrees F during emergency plant operation.
Additional investigation into this concern has revealed that low flows to the "A" control room A/C unit and RCIC room cooler also existed.
Engineering also determined that one ESF switchgear room cooler was originally undersized.
2.
Inadequate Cooler Nozzle Support a
On October 10, 1986, staff engineers of System Energy Resources, Inc.
(SERI, formerly Middle South Energy, Inc.) discovered that allowable nozzle loads at the inlet and outlet connections of cooling coils for 18 safety related room coolers supplied by American Air Filter (AAF) were not appropriately considered in the original piping design by the Architect Engineer, Bechtel Power Corporation. On October 16, 1986, SERI engineering in concert with Bechtel and AAF determined that nozzles on 4 of the 18 room coolers did not meet code allowables for upset and faulted conditions.
The nozzles on the other coolers were J16AECM87082601 - 5 gaa.. u.,
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mm determined to be acceptable based on a detailed stress analysis of connecting small bore piping and revised nozzle load interaction equations provided by AAF.
Subsequent analysis utilizing finite element methodology confirmed that the nozzles on 3 of the 4 coolers (which were initially identified as being overloaded) were adequately supported and met all code allowables and revised nozzle load interaction equations. The nozzles on the "A" cooler for the Fuel Pool Cooling and Cleanup (FPCC) pump room would have been overloaded during certain unlikely events such as an Operating Basis Earthquake (0BE) or a Safe Shutdown Earthquake (SSE). However, the cooler would have remained functional as the stress levels were significantly below the material ultimate strength.
Although the ECCS systems were not required to be operable when the condition was determined, the condition of the FPCC pump room cooler was not in agreement with the License Condition 2'.C(21).
This License Condition requires the Residual Heat Removal (RHR) system to be dedicated to the fuel pool cooling mode if " spent irradiated fuel is placed in the spent fuel pool prior to installation and operability of the safety related backup fuel pool cooling pump room coolers."
Because the "B" loop of SSW was out of service and the "A" FPCC pump room cooler may not have been fully functional due to the nozzle loading concerns, the RHR system should have been dedicated to the fuel pool cooling mode but was not due to licensed personnel failing to realize the linkage of the requirement.
This license condition was generated to document long term upgrade requirements for FPCC and was not directly associated with nozzle load concerns.
3.
Auxiliary Building Corridor Temperature Discrepancy As part of the overall effort to re-evaluate heat loads, water flows, and cooler performance in various rooms within the auxiliary building and the control room, a discrepancy was identified by SERI engineering staff regarding post-accident temperatures in some building corridors.
During accident conditions the auxiliary building was assumed by Bechtel in the original design to remain at the normal ambient conditions based on engineering judgement. As a result of efforts to confirm key assumptions, engineering analyses revealed that corridor temperatures would most likely exceed 104 degrees F, the maximum TS allowable temperature during normal operation.
Based on these latest conservative analyses, maximum corridor temperatures have been established. Temperatures resulting from the new analysis ranged from l
approximately 104 degrees F to approximately 121 degrees F.
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,8.e w s wim Further investigations into the impact of the increase corridor temperature, revealed that two room coolers were slightly deficient in capacity due to the increase heat loads, i.e., additional heat transfer from corridor into the rooms. The impacted room is the FPCC pump room. Since this condition of increased corridor temperature was
. associated with a LOCA while operating or in hot shutdown, the condition was not considered to affect operability in Modes 4 and 5.
A final assessment of the impact of this discrepancy revealed that i
the condition of increased corridor temperatures did not adversely I
affect equipment operability in any operational mode; therefore, plant restart was not constrained by this issue. Additional detail on those analyses performed as corrective actions is provided in Section E.3 of this report.
D.
APPARENT CAUSE 1.
Inadequate Cooler Flow Inspection end cleaning efforts revealed that the low flow conditions were due to the accumulation of material in varying degrees from the i
Preliminary investigative efforts identified that the SSW piping blockage and fouling was due to sedimentation in the piping common to PSW and SSW systems, i.e., material deposited from PSW.
However, a more comprehensive engineering assessment now attributes the system blockage to two primary root causes:
a)
Microbiological 1y induced corrosion (MIC) in the SSW basins is now recognized to have created increased pipe surface roughness and results in an increased resistance to flow.
b)
Sedimentation from PSW interfaces aggravates the MIC problem.
This was particularly in evidence in the control room A/C cooling water lines and the ESF switchgear room cooler lines.
Based on an engineering review of the issue, the following summarizes the mode of deposit formation for SSW piping systems shared with the PSW system.
In these piping systems the deposit builds up concentrically on the. pipe internal surface resulting in restricted flow and pitting under the deposits. The deposits are predominantly ferric iron with trace amounts of calcium, manganese, and phosphorous. The piping internal deposits show a " layered" appearance consistent with the general depositien characteristics of the iron oxidizing bacteria.
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l However, since these lines had continuous flow, slime-forming bacteria may have had a significant effect in assisting the deposit formation.
The slime-forming bacteria produce a bio-film on the internal surface of the pipe which tends to attract and hold the suspended solids contained in the water. These suspended solids then provide locations for the iron oxidizing bacteria to attach and deposit additional material (ferric iron resulting from the bacterial action).
In order to minimize this mode of corrosion / sedimentation in the future, the water treatment for the SSW system has been changed as discussed in the Section E.1.e.
It should be noted that the particular SSW piping system configurations for the "A" control room A/C unit and the RCIC pump room cooler may have contributed to the formation of_ deposits and low flow conditions.
Also low fluid velocities are believed to accelerate the overall rate at Which flow impeding materials are accumulated / deposited.
In addition to the reduced flow problem, the Division 2 ESF switchgear cooler at El.119' E was evaluated to have insufficient ~ heat removal capability under certain extreme conditions such that the cooling coil required replacement.
This deficient capacity was determined to be the result of an inadequate accounting by the A-E of heat transmission loads from surrounding areas, in particular the containment w.i under v
an accident condition.
2.
Inadequate Cooler Nozzle Support With regard to nozzle loadings, the nozzles of the coolers are connected to small bore piping. The support design for much of the small bore piping (2" and under) was performed by field personnel per the requirements of Mechanical Standard M-18.
M-18 is based on conservative stress limits and assumptions, and ensures that piping stresses will be kept well within the code allowable values without requiring a detailed, computer stress analysis to be performed.
l The conservative bases for the M-18 methodology ensures that loads imposed on equipment nozzles are acceptable for most types of equipment encountered.
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- - 0l2l9 0l 9 q7 0F 214 However, the AAF supplied room coolers are of light weight type design and are constructed of a low yield strength material.(e.g., copper-nickel) having minimal support within the 3mponent to limit bending at the nozzles. The light weight design plus highly conservative load-interaction equations provided by AAF resulted in extremely low nozzle allowable loads.- Other than this special type of equipment, there is no reason to believe other types of equipment would have allowable nozzle loads that are less than anticipated M-18 loads.
Furthermore, piping originally supported per M-18 guidelines results in very low stresses in the piping when reanalyzed by the ME101 4
computer' program. This provides confidence that loads imposed on equipment nozzles will be kept within reasonable ranges. This has been demonstrated by the acceptable ME101 analyses performed on the previously M-18 designed cooler piping.
In addition, a comprehensive review of small bore piping nozzle interface did not identify any other light weight components similar to the AAF supplied coolers.
Sample calculations from each category of heavy weight components.(such as tanks, pumps and heat exchangers) which typically have a rigid design were performed.
For all nozzles analyzed, nozzle loads were found well within the allowables.
Therefore it is concluded that the nozzle overload concern is limited to the uniquely light weight designed AAF supplied coolers.
3.
Auxiliary Building Corridor Temperature Discrepancy During accident conditions the auxiliary building was assumed by Bechtel the A-E to remain at the normal ambient conditions based on engineering judgement. More rigorous techniques utilizing conservative assumptions, particularly in the treatment of the containment heat source term, revealed that the original design assumption was not conservative.
E.
SUPPLEMENTAL CORRECTIVE ACTIONS 1.
Inadequate Cooler Flow a.
Hydrolazing/ Flushing Flows were improved to the coolers by hydrolazing and flushing the PSW/SSW piping to reduce the amount of blockage and by appropriate valve adjustments. When this work was completed some flow rates to the ESF electrical switchgear room coolers J16AECM87082601 - 9 2.....
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Con nol Room A/C 14 l! 4 125 136.8**
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ESF Switchgear Room Cooler-119'W 40
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ESF Switchgear Room Cooler 139'E 40 y,{23.5 38.2 5.
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ESF Switchgear Rooe, Cool _er 139'E 40
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Sys' tem Modifications Flush and drain connections have been installed in ESF switchgear rcom cooler piping to facilitate more routine future cleaning operations. Based on the nature of the PSW deposited material and
- - on experience gained during this outage, flushing of this piping should be ef.fective in controlling PSW related material deposition in the future.
Flushing and any other cleaning activities will be accomplished as determined by the performance monitoring program (SectionE.1.g.below).
The RCIC room cooler piping has been replaced with larger diameter piping to provide reduced resistance to flow. The coil of one Division 2 ESF switchgear room cooler located at E1.119' East has been replaced with a larger capacity cooling coil.and its fan speed has been increased.
Permanent flow monitoring instrumentation has also been installed in ESF switchgear room cooler piping to facilitate a periodic monitoring program. The monitoring program will be used to identify the need for piping flushes or other cleaning activities.
c.
Controls on Flow Monitoring Instrumentation To preclude the recurrence of using an incorrect flow coefficient for the flow devices used, annubar flow instruments will be controlled and issued as measuring and test equipment (M&TE) which will be stamped or have a metal tag attached with information which includes the flow coefficient and other pertinent data, d.
Maximization of Piping Flcw Since the deposition of PSW related material is apparently accelerated at low fluid velocities, steps have been taken to maximize these velocities.
Following the recent cleaning activities, system flow balancing was accomplished such that ficw to the affected coolers was maximized to the extent practical by manipulating discharge (return) throttle valves.
1 I
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Chemical Treatment To minimize flow degradation effects in SSW piping systems from microbiologically induced corrosion, an improved chemical treatment program has been established and implemented. The program utilizes a general corrosion inhibitor in conjunction with a biocide to minimize the microbiological activity and thus minimize the corrosion attack on the piping walls as well as the buildup of material that would tend to increase pressure losses and degrade flow.
In addition, a new water treatment program has been implen.ented for the PSW system. The new program utilizes a more effective dispersant and is expected to be more effective in minimizing or preventing fouling.
f.
Flow Monitoring Program As a long term corrective measure, a flow monitcring program has been established to provide flow performance and trending in. formation. The primary objective of tFe monitoring program is to i@ntify the need for flushing or cleaning those SSW components which are serviced by PSW during normal operating conditions. On a monthly basis, flow data will be measured and recorded for the ESF switchgear room coolers and the "A" control room A/C unit. The "B" control room A/C unit is excluded from the periodic monitoring due to its relatively high measured flow rate (in excess of 180 gpm).
if, however, significant deterioration in flow to the "A" control room A/C unit is observed in the monitoring program, the "B" side unit flow will be confirmed to be acceptable.
Flow thr sholds have been estelished to assure flow rates are maintained above the minimum design ilow values.
If measured flow is confirmed to be below this threshold, an evaluation will be performed to determine actions necessary to restore the flow or to increase monitoring to assure flow is maintained above minimum design flow.
This evaluation will include consideration of any trend noted and the potential for sudden further changes in cooler flow.
Should measured flow fall below the minimum design flow, the affected component will be considered inoperable. At the second refueling outage, one SSW division will be selected and each component's flow will be measured.
Furthermore, a review of the program's accumulated data will be performed. Based on this data and subsequent review, the need for additional flushing or cleaning will be determined. Threshold and monitoring periodicity will also be revised, if necessary, based on that evaluation.
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m.c,,,,asaw,m This program of cooler performance monitoring represents SERI's long term program and supersedes previous interim commitments to monitor the ESF switchgear room coolers and the control room A/C units (AECM-86/0283, 0309, and 0319).
The flow monitoring program has been implemented via plant procedure.
4 g.
Update to Design / Licensing Documentation The affected system flow diagrams will be revised to ~eflect the recently established minimum design flow values by December 31, 1986.
Appropriate changes to the Updated FSAR will be developed and incorporated into the December 1987 revision.
2.
Inadequate Cooler Nozzle Support Inspections of selected cooler coil tubing connections and header welds were performed to assist in the evaluation of the* nozzle loading concerns. No indication of excessive stresses was found.
Piping and pipe support modifications were completed on October 22, 1986 to resolve the nozzle load concerns. Training to Bechtel and NPE engineers involved in the piping stress analysis has been provided to reinforce the requirement that the vendor nozzle loading allowables be appropriately considered in the stress analysis.
Existing vendor nozzle loading allowables shall be utilized where available.
A standing order was issued on October 20, 1986 instructing each Goerations shift of the actions that should be taken during the time that b; FPCC room coolers were not operable. This action was taken in order to comply with License Condition 2.C(21).
3.
Auxiliary Building Corridor Temperature Discrepancy a.
Room Cooler Performance The impact of increased corridor temperatures on room cooler performance has been evaluated. With two exceptions, sufficient margin existed in cooler capacities such that post-accident design temperatures would not be exceeded.
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, e %,.n,,m The two components affected are the FPCC pump room coolers. The w ximum post-accident room temperature has been evaluated based on existing cooler capabilities and the revised heat loads. That temperature was determined to be less than 111 degrees F.
It has been determined that there are no 10CFR50.49 environmentally qualified components in this room.
Documentation for safety related equipment in the FPCC pump room (non-10CFR50.49) has been reviewed to confirm that the equipment's safety function will not be impacted by the increased temperature. As a result of this assessment, it was concluded that this increase in maximum room temperature was acceptable from a design and safety standpoint.
b.
Operator Activities in Corridor High Temperature Environment With conservative analysis predictions of excessive post-accident temperature in the auxiliary building corridor (op to approximately 121 degrees F), special measures have been formalized to insure that adequate protectiori is afforded the operators who may have to traverse or work in the corridor areas under these conditions. Several forms of insulated or air conditioned suits and equipment are ut'ilized at GGNS for the purposes of personnel protection against excessively high temperature environments.
Suits have been used and have provided protection in maintenance activities at GGNS in acbient temperatures in excess of 125 degrees F.
Appropriate administrative controls have been implemented to insure that equipment suitable for this purpose is available for personnel use in a post-accident environment.
c.
Control Building Heating and Air Conditioning Systems The impact of increased auxiliary building corridor temperatures on various safety related control building heating and air conditioning (HVAC) systems has been evaluated. This included the control room HVAC system and the safeguard switchgear and battery room ventilation system.
Based on this impact evaluation, it was determined that these systems' capability to fulfill post-accident safety functions as described in FSAR 9.4 was not adversely affected. No corrective measures with regard to these systems was necessary.
d.
Environmental Qualification of Corridor Equipment The impact of increased post-accident temperatures was evaluated for equipment in the auxiliary building corridor. Based on a review of i
/
GGNS equipment qualification documentation, it has been concluded l
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wame w me im that the increase in maximum corridor temperature for a LOCA/ loss of power event did not invalidate the qualification of components to function as required in the 100 day post-accident environment.
An evaluation has also been performed to determine the impact of the increased temperature on the affected equipment's qualified life.
Limiting components were identified to be Rosemount Model 1151 l
transmitters. These components were previously scheduled to be replaced during the next refueling outage (end of 1987) due to the previously established expiration of qualified life. An analysis of the increased thermal aging affects from a higher post-accident corridor temperature has confirmed these limiting components need not be replaced any earlier than the current schedule, i.e., the second refueling outage.
i In conclusion, the increased corridor temperature has no impact on the qualification of corridor equipment as required by 10 CFR 50.49.
Further, while revisions a engineering environmental equipment documentation and revisions to certain components qualified life are necessary, no actions to replace equipment prior to t.1e current schedule are required.
e.
Documentation Update / Revision From an equipment qualification standpoint, no immediate corrective actions are required with regard to equipment as described in the next section. However, the increased temperatures in the auxiliary building corridors and FPCC pump room will require changes in the engineering documents describing the post-accident environments and components' qualified lifetime. These documents (Engineering Specification E100 and Electrical Standard ES19, respectively) have been updated to reflect the impact of this increased maximum post-accident corridor temperature.
f.
Additional Design Review of SSW System As a result of discrepancies identified in the design assumptions 4
made by Bechtel and other apparent original design discrepancies
)
identified in LER 86-029, the SERI Independent Safety Engineering Group initiated an "SSW System Performance Assessment". The purpose of this assessment was to (1) ensure the system design and reliability meet the requirements for safety related service and (2) identify enhancements which may improve. operation, reliability and availability of the system. An engineering design firm was retained to perform i
l.
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-- e w w nn an indepth evaluation of the original Bechtel design of the system. A multi-discipline cross-section of the design process was selected to assess the design's conformance to industry standards and regulatory requirements, adequacy of design flowrates, piping and support analysis, environmental qualification, materials of construction, Appendix R review, system instrumentation and electrical interface.
As discussed at the 2/4/87 enforcement conference, a second system was included in the final design assessment report.
Since many of the items evaluated for SSW are generic to the design process, the second system was selected to evaluate another unique aspect of the design process, the General Electric /Bechtel interface.
For this purpose the High Pressure Core Spray (HPCS) System was selected. The HPCS review was of a lesser scale than the SSW system assessment with the primary focus being on assessing the implementation of design criteria.
~
The SSW system design assessment, including additional SSW system interface evaluations and the HPCS review are complete and the assessment resolutions reported in AECM-87/0095 dated May 8, 1987.
The SSW System design assessment identified that a potential water hammer exists in the unlikely event that a LOP, LOCA, or LOP /LOCA occurs while the SSW system is in operation and the drywell purge compressor i-.ation valves or FPCC heat exchanger isolation valves are open. Specifically, partial draining of the drywell purge compressor and/or FPCC heat exchanger supply and return lines through open component isolation valves could occur if, during periodic surveillance testing, a LOP and/or a LOCA occurred resulting in SSW pump and component isolation valve power shedding. Approximately 20 seconds following pump shedding, pump sequencing with open component isolation valves could generate a water hammer by rapidly refilling the voids created during drainage.
As resolution of the pountial water hammer concern, the SSW system control logic will be modified to support drywell purge compressor and FPCC heat exchanger isolation valve closure prior to SSW pump restart. Since only these components are susceptible to drainage and the component isolation valves are located below the maximum drain down elevation of the system, any voids created during drainage will be isolated prior to pump restart. The SSW system functional design requirements also necessitate that orificed bypass lines be installed to gradually refill the voids in the isolated supply and return lines following pump restart.
Due to procurement schedule uncertainties concerning the receipt of additional equipment, the design l
modifications may not be complete during the sccond refueling outage j
but will be complete prior to the third refueling outage.
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SAFETY ASSESSMENT
1.
This section provides a safety assessment of various cooler flow and heat load related issues identified at GGNS starting with the fall outage in 1985.
Key events are listed below in chronological order with a brief discussion of each. The discovery in November 1986 of inadequate engineering assumptions regarding the auxiliary building corridor temperature required a re-evaluation of the key deportability determinations made during and since the fall 1985 outage. A brief discussion of those re-evaluations is also provided with event summary.
This section also provides a safety assessment of the inadequate cooler nozzle support concern identified on several coolers.
2.
Key Event Chronology and Safety Assessment a.
November, December 1985 1)
Flow balancing of the SSW "B" loop following the installation of a larger capacity SSW pump impeller resulted in the identification of low flow in certain ESF switchgear room coolers. This discovery led to flow measurements in SSW "A" and the identification of similar low flow conditions in SSW "A".
As a result the affected ESF switchgear room coolers were cleaned or replaced.
2)
It should be noted that since SSW "B" flow data was obtained following impeller replacement, cooler flow information for SSW "8" prior to the commencement of the fal? 1985 outage was not available. An assessment of the most appropriate, available flow information for SSW "A" and "B" revealed that the affected coolers were able to maintain room temperatures within environmentally qualified equipment temperature requirements for operability. Therefore, the condition was determined at that time to be not reportable.
3)
Increased post-accident auxiliary building corridor temperatures required a re-evaluation of the above safety assessment. The December 1986 re-evaluation concluded that the peak post-accident temperature which would be reached in an ESF switchgear room is 144 degrees F.
This peak temperature is within the requirements for environmentally qualified equipment in the affected room, and it was concluded that the equipment would have remained operable to accomplish its intended safety function. 6Jditional detail on i
equipment qualification considerations is provided in Item 2.b below.
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conclusions on deportability from the fall 1985 outage were unchanged. This re-evaluation and conclusion included consideration of the affects of incorrect flow measurements which occurred during that outage.
b.
August 1986 1)
The use of incorrect annubar coefficients during the fall 1985 outage was identified.
%rther, it was determined that this deficiency impacted certain components' flow measurements performed on both SSW divisions in November and December 1985. As a result of the discovery, additional flow data for affected components was taken and determined to be less than nominal design values.
Coolers were flushed to the extent practical. An assessment of the as-found flow rates determined that the condition was reportable based on a single condition causing flows in' both SSW divisions to be less than nominal design values. LER 86-029-00 was submitted.
l Based on conditions identified at the time, an evaluation was
{
performed. The most limiting component was determined to be able l
1 to withstand a 140 degree F environment for the required 100-day post-LOCA period. The analysis concluded that all rooms except two l
could be maintained at a temperature of less than 140 degrees F.
j One of the ESF electrical switchgear rooms was analyzed at temperatures being maintained below 141 degrees F and the other
{
below 147 degrees F.
No credit for the coolers ability to remove
)
heat was taken for these two rooms.
2)
The most limiting components were the 7.2 KV power breakers which provide power to the reactor recirculation pump motors. These breakers, however, would complete their required safety function well before reaching the 140 degree F temperature during postulated l
accident conditions. The second most limiting components were J
relays installed in ESF load centers.
Calculations show that these components could operate in a 158 degree F environment for at least 100 days. This evaluation concluded that the affected equipment would have remained operable to accomplish its intended safety function.
3)
To assess the impact of increased post-accident auxiliary building heat loads, a re-evaluation of the peak temperatures was performed for SSW "A" and "B" ESF switchgear rooms.
Revised room heat loads were developed accounting for the corridor temperature increase and several other factors. Using revised heat loca and "as-found" ESF switchgear room cooler flew values, peak post-accident temperatures I
I i
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4)
Due to the sequence of cleaning and flow measurement, "as-found" data was not available for certain components, including the RCIC pump room cooler and the SSW "A" Control Room A/C unit. These components are discussed in item 2.c below.
c.
September, October 1986 1)
The GGNS first refueling outage commenced in early September, 1986.
Subsequent flow measurements revealed low flow condition in the "A" control room A/C unit and RIIC pump room cooler.
Discretionary enforcement was requested and granted to permit RF01 outage activities, i.e. modes 4 and 5.
SERI letters AECM-86/1283 and 86/0319 provided safety assessment for those activities.
Numerous nonconformance reports were generated to track the conditions identified.
As a basis for discretionary enforcement and for continuing with outage activities in a safe manner, minimum acceptable flow rate values were established for outage conditions such that design temperatures would be maintained in the affected areas. A periodic flow monitoring routine was implemented to confirm that actual flow rates met or exceeded the minimum flow criteria during the outage.
2)
RCIC pump room heat loads were adjusted to account for the increased corridor post-accident temperature. Taking no credit for cooler operation, the peak post-accident temperature was determined to be 154 degrees F.
The limiting 10CFR50.49 component type in areas serviced by the RCIC room cooler was determined to be ASCO solenoids Further evaluation confirmed that the increased temperature levels l
would have had no impact on RCIC operability throughuut the analyzed event.
It should also be noted that no credit is taken for RCIC system operation in the mitigation of the loss of coolant accident which is generating the containment heat source term in this scenario; thus, the safety significance of the low flow identified in the RCIC pump room cooler is further reduced.
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With regard to the "A" Control Room A/C unit, the minimum measured flow was 120 gpm.
Based on this condenser flow, the unit was evaluated to have sufficient capacity to maintain control room temperature below 72 degrees F when in the recirculation mode.
Fresh air makeup to the control room is required following the initial 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> of isolation post-accident.
Introducing outside air under conservative assumptions would increase the control room temperature (with a condenser flow at 120 gpm).
Calculations of unit capacity have determined that control room temperatures w1il rise during the periods of fresh air intake.
However, the temperature was determined not to exceed the upper limit of human factors comfort zone (77 degrees F as discussed in AECM-86/0319).
Peak control room temperature was therefore determined to remain below the 90 degrees F limits based on equipment qualification consideratic s.
d.
November 1986 1)
Inadequate engineering assumptions in the original design regarding auxiliary building corridor temperature were identified. The increased temperature required re-evaluation of room heat loads and corridor equipment qualification.
Corrective actions necessary to permit plant restart were t
identified and taken.
2)
The corrective actions included an appraisal of the impact of the increased heat loads on auxiliary building room cooler performance. As discussed in Section E.3.a of this report, the increased heat loads are either within current cooler capabilities or the resulting increased room temperature is acceptable from a design and safety standpoint (FPCC pump room).
Impact on equipment qualification was also evaluated for equipment in the auxiliary building corridor. As concluded in Section E.3.a of this report, the increased corridor temperature represents no impact on the qualification of equipment as required by 10CFR50.49.
3.
Inadequate Cooler Nozzle Support With regard to nozzles on the "A" cooler for the FPCC pump room, a detailed stress analysis was performed. The code allowable limits l
and NRC provided operability criteria were exceeded at only one location in the tube wall.
Further evaluation was made utilizing
)
i elastic-plastic methodology and maximum loads calculated from the finite element model of the cooler.
It was concluded that even though I
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elastic limits (0.2% strain) are exceeded, plastic deformation stopped 3t 2.05% strain. Even though the strain at failure is 42% for this copper-nickel material (SB-111, annealed), a value of 28% was conservatively used in the analysis. Thus, for the faulted loading there is a margin of safety of 13.7 (28/2.05). Based on critical buckling, a margin of safety of 5.5 (866 in-lb/157 in-lb) exists for the tube wall. Therefore, it is concluded that there will be no significant deformation to restrict flow or cause leakage, and operability of the subject cooler will be maintained even in the enlikely event of SSE and LOCA occurring simultaneously.
G.
This section has been generated in order to identify design deficiencies found in the SSW system during a special design review previously addressed in Section E of this report.
1.
REPORTABLE OCCURRENCE During a special design review of the Standby Service Water (SSW) system, a design deficiency was identified.
During LOCA conditions coincident with a loss of offsite power (LOP), a single' failure of either of two motor control centers could cause an SSW flow diversion resulting in a gradual reduction of the 30-day Ultimate Heat Sink (UHS) inventory through the nonessential Plant Service Water (PSW) System.
The review also identified two additional design concerns which could similarly jeopardize the UHS inventory, if the system was initially lined up to utilize special design features. These conditions are reported in accordance with 10CFR50.73(a)(2)(vii) and 10CFR50.73(a)(2)(v).
2.
INITIAL CONDITIONS Each design deficiency was identified while the plant was operating between 85 and 100 percent power.
3.
DESCRIPTION OF OCCURRENCE An SSW system performance essessment was performed by the SER' Independent Safety Engineering Group. During this review design concerns were identified which could jeopardize the 30-day water inventory of the SSW/VHS system.
The SSW system provides cooling water to plant auxiliaries that are essential for a safe reactor shutdown and also serves as the UHS during accident conditions. When the plant is operating, the SSW system is normally in standby, and cocling water to certain components is supp'ied by connected nonessential cooling water systems.
Per FSAR 9.2.1.2, redundant automatic isolation valves are provided to separate all nonessential cooling water systems from the SSW system.
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=w.,im the design review revealed a failure to provide suitable redundancy such that failure of an MCC (rather than complete failure of a division of ESF power) could create a path for the diversion of SSW flow such that a gradual reduction in the 30-day UHS inventory could occur.
a.
A design deficiency identified on March 18, 1987, during the special review pertains to the PSW/SSW interface for the Control Room air conditioning istem. The Control Room air conditioning units are aligned to e ; PSW system for cooling during normal plant operation. During LOP or LOCA conditions, the cooling water supF.y automatically transfers to SSW. The PSW/SSW interface isolation valves (GG-1KG-20-FOXX) are not suitably redundant, since some of the valve motor operators share a common power supply (GG-1ED-MCC-XXXX). This deficiency is common to both divisions of SSW. Under LOCA conditions concurrent with a loss of offsite power, the single failure of either motor control center (MCC) 15B61 or MCC 16B61 would align the affected 35W loop to the PSW system via a 3-inch header which would result in the gradual loss of UHS inventory through the-PSW system.
A simplified drawing of the affected portion of the system is attached for reference.
Under LOP or LOCA condit S supply valves F125 and F066A supply valve F064A receives a receive a signal to close ' v
,, a n signal to open, aligning SSW flow t'nrough the "A" Control Room air conditioning unit. The motor operators for valves F125 and F066A are powered from the same motor control center, MCC 15B61, while valve F064A is powered from MCC 15B21.
If MCC 15B61 failed prior to the automatic transfer from PSW to SSW during a LOP or LOCA situation, PSW supply valves F125 and F066A would remain open due to the loss of power to their motor operators.
SSW supply valve F064A would automatically open as designed because of its separate power supply. Thus, a single failure associated with MCC 15861 during LOP /LOCA conditions could align the SSW "A" supply to PSW with a resulting reduction of UHS system inventory. Likewise, a single failure of MCC 16B61 could align the SSW "B" discharge from the "B" Control Room air conditioning unit to PSW through valves F074B and F189. This would result in diverting a portion of SSW through the PSW discharge from the "B" Control Room air conditioning unit.
Upon notification of the finding at 1500 on March 1<8, Operators entered into a Limiting Condition for Operation (LCO) for the SSW system. By 1530 the SSW system was restored to operable status by isolating and tagging out the affected PSW valves and manually starting the SSW system to supply cooling water to the Control Room air conditioning units.
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The system. review also discovered that the two SSW basin blowdown vsives on each SSW loop are powered by common MCCs.
If a LOCA occurred while an SSW loop was operating in the blowdown configuration (pump discharge lined-up to the discharge basin), a single failure of the MCC would cause the valves to remain open which could jeopardize the UHS inventory if no operator action is taken to stop the discharge. This event is less credible since SSW blowdown is normally accomplished while the reactor is shutdown.
In the past 3 years SSW blowdown was performed only 3 times while the reactor was operating. Of these, the longest single blowdcun lasted 3.7. hours. Actions have been taken to preclude the use of the blowdown function during normal operation until further investigations can be performed.
c.
It was also determin'd that the two isolation valves on the "B" SSW loop which serve as a backup supply of c'coling water to the instrument air a'J service air compressors are powered from a common MCC and also share a common relay in their automatic closure circuit. Normally, the Turbine Building Cooling Water (TBCW)
System supplies cooling water to the instrument air and service air compressors. The SSW system may be manually aligned to supply cooling water to the compressors. The cooling water piping associated with the compressors is non-safety grade.
If SSW were aligned to the compressors prior to a LOCA, a failure of either the MCC or the common relay would prevent automatic isolation of this nonessential portion of ti'e system. In order for this situation to occur, the system must be manually aligned to provide air compressor cooling water from the SSW system. An automatic transfer is prohibited during a LOCA but will oc<or on a LOP, if a LOCA signal is not present. A loss of offsite power and automatic transfer to SSW cooling follored by a LOCA and subsequent failure of the MCC or relay combined with a breech of the air compressor cooling water piping is considered to 'e outside of the plant design basis.
v 4.
APPARENT CAUSE The cause of the concerns was determi"ei to be a failure of the architect engineer, Bechtel, to prov Q uitable isolation capability for these specific SSW system interfaces.
A previous situation involving a potential for having an VHS inventory less than 30 days was reported in LER 84-031. The cause of that situation was due to the system dependence on nonoperational Unit 2 equipment to maintain the 30-day supply. A siphon line was installed between the two SSW basins to ensure a 30-day UHS inventory is available to either loop of SSW. The 1984 event and the concerns discussed in this report are related but are not considered to be similar in nature.
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e % m mm A second incident involving a partial loss of SSW cooling basin inventory was reported in LER 83-075. The cause was due to a system design which would allow the SSW basin water inventory to be siphoned back out of the basins through PSW piping if a LOP occurred during the basin makeup process. The 1983 event and the concerns described in this report are considered similar in nature since they address loss of VHS water inventory through the PSW system during a LOP.
5.
SUPPLEMENTAL CORRECTIVE ACTIONS Administrative controls have been implemented to ensure that operator action is taken within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> to isolate the PSW flowpath under the LOP /LOCA conditions with a power failure of either MCC 15B61 or MCC 16B61. With these controls in effect, the SSW system was returned to standby, and the PSW system flow was restored to both Control Room air conditioning units. As a permanent resolution to' provide suitable isolation capability, the PSW supply valve (QSP41F125-A) and the PSW return valve (QSP41F189-B) will be powered from MCCs other than MCCs 15B61 and 16B61, respectively.
To prevent potential loss of UHS inventory during blowdown operations, permanent administrative controls have been placed in plant procedures to require that an operator be stationed at the SSW blowdown line isolation valves when blowdown is in progress.
SERI has evaluated the SSW supply to the air compressors.
It was determined that the leakage from a postulated moderate energy pipe crack in the non-essential portion of the system would not substantially reduce system flow rates or result in a reduction of the 30-day UHS inventory.
Therefore, suitable isolation exists at this interface and no design modifications or administrative controls are required.
The SSW system design assessment, including additional SSW system interface evaluations, are complete and the assessment resolutions reported in AECM-87/0095 dated May 8, 1987.
6.
SAFETY ASSESSMENT
a.
If a failure of either MCC 15B61 or MCC 16861 occurred prior to an automatic transfer of PSW to SSW during a LOP /LOCA, the UHS 30-day inventory would be reduced if no operator actions were taken to stop the diversion of SSW flow through the PSW interface piping.
Assuming worst case postulated conditions, operator action is required within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> to ensure the 30-day UHS inventory is
(
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.uwim maintained. Administrative controls were implemented to ensure that operator action is taken within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> to isolate the PSW flowpath under the LOP /LOCA conditions with a power failure of either MCC 15B61 or MCC 16B61.
It should be noted that failure of either MCC following proper realignment of the PSW and SSW valves as a result of a LOCA signal car.not cause a flow diversion.
If only a LOCA occurred with offsite power available, the PSW system could be operated to supply makeup to the SSW basin to maintain the basin level if required.
b.
The postulated scenario discussed under item 3.b could conceivably reduce the available VHS 30-day inventory; however, I
the SSW system !s not normally configured to utilize this feature. Administrative controls have been implemented to ensure an operator is stationed at the SSW blowdown valves during blowdown opentions, c.
Although the postulated scenario discussed under item 3.c could conceivably reduce the a/ailable VHS inventory, it was determined that this scenerio does not affect the available 30-day inventory and that suitable isolation exists at this interface.
d.
Because the UHS-inventory consists of the combined inventory of the two cooling tower basins, certain single failures (which would otherwise not be significant because of the independent, redundant SSW cooling loops) can impact the safety function of the unaffected SSW train due to the gradual reduction of the shared UHS inventory.
These essential /non-essential SSW interfece problems are not believed to be generic due to the uniqueness of this shared inventory. The cooling capability of the SSW system is not affected since both trains of SSW are operable in these scenarios.
Consequently, all cooling requirements are met by the unaffected
- train, l
J16AECM87082601 - 25 g....
Attachment to AECM-87/0169 I
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REEUUNCEE, INC.
Otr4p D Mt61Ev.JR E$$$ dens August 28, 1987 U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Attention: Document Control Desk Gentlemen:
SUBJECT: Grand Gulf Nuclear Station Unit 1 Docket No. 50-416 License No. NPF-29 Pipe Fouling Results in Low Flows to ESF Room Coolers and Related SSW Design Concerns LER 86-029-09 AECM-87/0169 Attached is Licensee Event Report (LER) 86-029-09 which is a final report.
Your r ly, s
00K:bms Attachment
/
cc:
Mr. T. H. Cloninger (w/a)
Mr. R. B. McGehee (w/a)
Mr. N. S. Reynolds (w/a)
Mr. H. L. Thomas (w/o)
Mr. R. C. Butcher (w/a)
Dr. J. Nelson Grace, Regional Administrator (w/a)
U. S. Nuclear Regulatory Commission Region II 101 Mariette St., N.
,W., Suite 2900 Atlanta, Georgia 30323 Mr. L..L. Kintner, Project Manager (w/a)
Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20814 i
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