ML17229A212
| ML17229A212 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 01/28/1997 |
| From: | Stall J FLORIDA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-96-06, GL-96-6, L-97-18, NUDOCS 9702040264 | |
| Download: ML17229A212 (30) | |
Text
CATEGORY 1 REGULATOR INFORMATION DISTRIBUTIONOSTEM (RIDE)
ACCESSION<NBA'9702040264 DOC.DATE: 97/01/28 NOTARIZED: YES FACIL:50<<335 Gt. Lucie Plant, Unit 1, Florida Power
& Light Co.
50-389 St. Lucie Plant, Unit 2, Florida Power
& Light Co.
AUTH.NAME AUTHOR AFFILIATION STALL,J.A.
Flori'da Power
& Light Co.
RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
DOCKET 05000335 05000389
SUBJECT:
Submits 120-day response to GL 96-06, "Assurance of Equipment Operability
& Containment Integrity During Design-Basis Accident Conditions."
Summary Rept encl.
DISTRIBUTION CODE:
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SIZE: 20 TITLE: GL 96-06, "Assurance of Equip Oprblty
& Contain.Integ.
during Design NOTES:
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Florida Power &Light Company, P.O. Box 128, Fort Pierce, FL 34954-0128 January 28, 1997 L-97-18 10 CFR 50.4 10 CFR50. 54 (f)
U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 RE:
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 120-Day Response Generic Letter 96-06 This letter provides the Florida Power and Light Company (FPL) 120-day response to Generic Letter (GL) 96-06, Assurance ofEquipment Operability and Containment Integrity During Design-Basis Accident Conditions, for St. Lucie Units 1 and 2. By letter, L-96-280, dated October 28, 1996, FPL committed to implement the GL actions and submit the requested summary report by January 28, 1997.
The summary report is attached.
The GL requested that within 120 days of September 30, 1996, FPL submit a written summary report stating the actions taken in response to the GL, the conclusions that were reached relative to susceptibility for waterhammer and two-phase flow in the containment air cooler cooling water system and overpressurization of piping that penetrate containment, the basis for continued operability of the affected systems and components, and the corrective actions implemented or that are planned to be implemented. Ifsystems were found to be susceptible to the conditions that were discussed in the generic letter, FPL was to identify the systems affected and describe the specific circumstances involved.
FPL did not identify any piping susceptible to waterhammer or two-phase flow in the containment air cooler cooling water system.
The evaluation did identify piping sections for containment penetrations and pipe sections inside containment that could be vulnerable to thermal overpressurization.
Each of the identified cases which provide containment integrity or other safety related functions were analyzed for functional operability using the basis of the acceptance criteria contained in ASME Section IIIAppendix F as provided for within NRC Generic Letter 91-18.
The results indicate that all affected piping and valves within both units will remain functional with no violaton of the pressure retaining boundary under accident conditions.
FPL considers the operability assessments contained in the attached summary report to be valid for the current operating cycle of each unit and ifnecessary, due to NRC review schedules, the next operating cycle of each unit.
97020402b4 970i28 PDR'~"ADQCK 05000335 P
St. Lucie Units 1 and 2 Docket Nos, 50-335 and 50-389 L-97-18 Attachment Page 2 Section 7.0, Corrective Actions, of the attached summary report contains new regulatory commitments which are summarized below:
FPL willsubmit license amendment requests to incorporate ASME Section III, Appendix F, into the licensing bases of St. Lucie Units 1 and 2 by April30, 1997.
2.
Should FPL, during the preparation of the proposed license amendment, identify any of the Unit 1 piping configurations subject to thermal overpressurization that do not meet the criteria of ASME Section IIIAppendix F, FPL will install engineered corrective actions during the upcoming Fall 1997 refueling/steam generator replacement outage (SL1-15).
3.
For Unit 2, the next refueling outage (SL2-10) is scheduled to start on April 14, 1997.
Due to the extent of the required evaluation and the limited time remaining prior to the outage, there is insufficient time to complete the detailed engineering, procure materials and plan for the installation of any corrective actions during the upcoming Unit 2 outage.
FPL proposes to delay any corrective actions on Unit 2 until at least the next refueling outage (SL2-11), which is currently scheduled for the Fall 1998.
This letter and the attached summary report are provided pursuant to the requirements of Section 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f).
Please contact us ifthere are any questions.
Very truly yours, J. A, Stall Vice President St. Lucie Plant JAS/GRM Attachment cc:
Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, St. Lucie Plant
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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-18 Attachment Page 3 STATE OF FLORIDA
)
)
COUNTY OF ST. LUCIE
)
Ss.
J. A. Stall being first duly sworn, deposes and says:
That he is Vice President, St. Lucie Plant, for the Nuclear Division of Florida Power & Light Company, the Licensee herein; That he has executed the foregoing document; that the statements made in this document are true and correct to the best of his knowledge, information and belief, and that he is authorized to execute the document on behalf of said Licensee.
J. A. Stall STATE OF FLORIDA COUNTYOF W. LAC.H=
Sworn to and subscribed before me this 2,g day of
,19 9 by J. A. Stall, who is personally own to me.
Name of Notary Public - State of Florida KAREN WEST
~a'Y COit IIt8SSON ICC359928 EKPIRES AprN 18, 1998
'4~ '14" 8OtOHtllstjTROYMlINSURICE, PC.
KA-ErN lN~S T (Print, type or stamp Commissioned Name of Notary Public)
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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-18 Attachment Page 1
NRC Generic Letter 96-06 Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions Summary Report St. Lucie Nuclear Plant Units 1 and 2 Florida Power and Light Company
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-18 Attachment Page 2 TABLEOF CONTENTS 1.0 Abstract..................................
2.0 Description................... ~............
3.0 Methodology................................
Waterhammer Two-Phase Flow And CCW Pump NPSH Evaluations Overpressurization of Isolated Piping Sections
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4 4.0 Evaluation....................................
Waterhammer Evaluation Two-Phase Flow And CCW Pump NPSH Evaluation Overpressurization of Isolated Piping Evaluation Consideration of Single Failures 8
5.0 Operability Assessment Review Waterhammer Two-Phase Flow in Safety-Related Piping and Components Overpressurization of Isolated Piping Sections 13 6.0 Results Summary 14 7.0 Corrective Actions........................... ~...............
15 Table 1 - St. Lucie Unit 1 Penetrations and Piping Vulnerable to Thermal Overpressurization
. 16 Table 2 - St. Lucie Unit 2 Penetrations and Piping Vulnerable to Thermal Overpressurization, 17
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-18 Attachment Page 3 1.0 Abstract This summary report provides the Florida Power and Light Company (FPL) response to NRC Generic Letter (GL) 96-06 as it relates to St. Lucie Units 1 and 2. GL 96-06 concerns assurance of equipment operability and containment integrity during design-basis accident conditions.
GL 96-06 requested utilities to determine if containment air cooler cooling water systems are susceptible to either waterhammer or two phase flow conditions during postulated accident conditions and whether piping systems that penetrate the containment are susceptible to thermal expansion of fluid so that overpressurization of piping could occur.
An evaluation of the loss of coolant (LOCA) and main steam line break (MSLB) events for each unit with a coincident loss ofoffsite power (LOOP) indicates that voiding is expected to occur in a 16.1+
second time frame for the limiting case.
Considering the non-mechanistic nature of the pipe rupture assumed for the LOCA containment analysis, provision of large amounts of conservatism within the containment fan cooler (CFC) model is not warranted.
Elimination of CFC model conservatism for the limiting Unit 1 LOCA analysis increases the time frame before voiding is expected to occur to between 16.6-17.6 seconds.
As the emergency diesel generator (EDG) loading sequence places the component cooling water (CCW) pumps in service within 16 seconds after LOOP, voiding is not anticipated to occur for these scenarios for either unit.
For the limiting Unit 1 LOCA analysis, a review of the post-transient recovery phase indicates that pump net positive suction head (NPSH) willbe maintained and that two-phase flow conditions willnot occur within the CCW system.
As this analysis is bounding, voiding or two phase flow conditions will also not occur after system pressurization during the quasi-steady state design condition.
The evaluation identified pipe sections for containment penetrations and piping sections inside the Unit 1 and Unit 2 containments that are vulnerable to thermal overpressurization during LOCA and MSLB events.
Each of the identified cases that perform containment integrity or other safety related functions were analyzed for functional operability using acceptance criteria of ASME Section III, Appendix F, as provided by NRC Generic Letter 91-18. The analysis indicated that all affected piping and valves would remain functional under the accident conditions associated with NRC Generic Letter 96-06. Although the affected piping has been reviewed for functionality, more detailed analysis and evaluation are required to establish and demonstrate acceptance limitations within ASME Section III, Appendix F, as a long term basis for design.
The evaluation concluded that the affected St. Lucie Units 1 and 2 systems remain operable with respect to the design challenges addressed by Generic Letter 96-06.
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-18 Attachment Page 4 2.0 Description This generic letter requested evaluation of three main issues:
1.
Effect of hydrodynamic forces on cooling water systems serving containment coolers during LOCA and MSLB events; 2.
Two-phase flow conditions within cooling water systems serving containment air coolers as a result of heating during LOCA and MSLB events; 3.
Thermally induced overpressurization of isolated water-filled piping sections.
This report summarizes the evaluation performed by FPL to address the issues of Generic Letter 96-06 as they relate to St. Lucie Units 1 and 2. The Code of Record for St. Lucie Unit 1 is ANSI B31.7, 1969 Edition. The Code of Record for St. Lucie Unit 2 is ASME Section III, 1971 to 1973 Summer Edition.
3.0 Methodology Waterhammer The waterhammer evaluation determined whether the CCW system water would flash to steam in the containment fan cooler (CFC) cooling coils during design basis accidents coincident with a loss ofoffsite power (LOOP). Ifoffsite power is lost during a design basis accident, cooling water flow to the CFCs would be lost until the CCW pumps are restarted with electrical power supplied by the emergency diesel generators (EDG). During this period, containment temperature increases above the saturation temperature of the stagnant CCW inside the CFCs and the CCW inventory may boil unless cooling flow is reestablished within a suitable time frame.
Of interest are the conditions that would be generated within the CCW system by the thermal/hydraulic response of containment fan coolers during design basis accidents in conjunction with a LOOP. The rapidly heating containment atmosphere heats the CFC fins and tube metal and the component cooling water in the CFC tubes. Ifthe water in the CFC tubes reaches saturation temperature, steam will begin to form and voiding willoccur.
Should a void occur, potential waterhammer effects due to void collapse during the repressurization would need to be addressed.
Sargent &Lundy was contracted to assist FPL in responding to this issue by developing a calculation that would address the CFC heatup transient for the St. Lucie Units. The calculation is based on a computer model of the event which simultaneously solves the heat transfer and fluid flow equations governing the transient CFC heat-up mechanism.
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-18 Attachment Page 5 The model assumes that coincident with the initiation of a postulated MSLB or LOCA, a loss of offsite power (LOOP) occurs.
For the purposes of the evaluation, the coastdown time for the CFC fans is assumed to be much longer than the transients being evaluated.
Coastdown time for the cessation of CCW flow within the system is modeled as 2 seconds.
In evaluating the hydraulic transient, CCW flow is assumed to restart within 16 seconds after the LOOP. This timeframe is based on the EDG design basis, which is to start and accelerate to full speed within 10 seconds, and the CCW pump loading within 6 seconds of the EDGs reaching full speed.
Heat transfer to the metal of the CFC tubes and fins is determined from the CFC outside surface area and temperature, the fin efficiency, the post-accident temperature and humidity. Since saturated entry conditions are characteristic of the LOCA containment response, condensation dominates the heat transfer. For MSLB and LOCA cases, the condensation heat transfer coefficient for containment atmosphere in contact with components and structures inside the containment is taken as four times the Uchida correlations.
This approach is consistent with the Uchida's original data and is endorsed by NRC NUREG-0588, Standard Review Plan Section 6.2.1.5, and Branch Technical Position CSB-1.
The effect of convective heat transfer on heat gain becomes significant during an MSLB due to the high dry bulb temperatures for this event.
For MSLB cases, single phase forced convection coefficients for the CFC tubes were also calculated based on the Hilpert correlation.
For these MSLB
- cases, the total heat transfer to the CFC tube is conservatively calculated as the sum of the heat transferred using the external heat transfer coefficient based on condensation only, and then based on convection only.
The heat transfer to the water inside the CFCs is determined from the inside surface area and temperature of the CFC tubes, the water temperature and pressure, and the forced convection or free convection coefficient (whichever is greater) or nucleate boiling heat transfer coefficient at the inside of the CFC tubes.
The Dittus-Boelter correlation is used for single-phase forced convection, the Catton correlation is used for single-phase free convection, and the Rohsenow correlation is used for nucleate boiling.
Heat transfer from the containment atmosphere to water-filled return line piping is determined from the pipe outside surface area, the post-accident temperature and the condensing heat transfer coefficient (four times the Uchida correlation).
Heat transfer from the return line piping to the water inside it is determined from the pipe inside surface area and temperature, the water temperature, and the larger of the single-phase forced convection coefficient (Dittus-Boelter correlation) or single-phase free convection coefficient (Catton correlation).
St, Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-18 Attachment Page 6 The difference between the heat transferred to the metal and the heat transferred to the water is the rate of heat addition to the CFC tubes and fins; this results in the metal temperature increases.
The model includes a heat transfer correlation to evaluate forced convection condensation internal to the return piping and a hydraulic model of the shortest piping path to the surge tank (return path friction, volume and momentum) in order to evaluate the affects of any steam voiding within the system.
This aspect of the model was not utilized as steam voiding did not occur in the timeframe of interest.
In evaluating the approach to CCW saturation temperature, the pressure of the CCW in the coolers is of interest.
The minimum pressure occurs during stagnant conditions.
During flow conditions, the pressure at the CFC coolers is higher (due to pump operation and flow restrictions in the CFC return lines).
In both units, the elevation of the top row of tubes in the HVS-1C cooler (72.3 ft and 71.4 ft, respectively) is higher than the other three coolers (which are approximately 17 to 21 feet lower). Due to the difference in static head, water in the HVS-1C cooler will reach saturation temperature before it will in the other coolers.
2.
The surge tank low level alarm setpoint elevation is 75.67 feet in Unit 1 and 74.92 feet in Unit 2.
3.
The difference between the low level alarm setpoint elevation of the surge tank and the elevations of the HVS-1C coolers leads to a CCW pressure of 16.2 psia (saturation temperature 217F) in the coolers under stagnant flow conditions.
The model focuses on the heatup of the CCW within the CFC coolers rather than the heatup of the water within the CCW piping.
Heatup within the coolers will occur more rapidly due to their construction (higher thermal conductivity, thinner wall, greater area to volume ratio due to the extended heat transfer area).
Four cases were analyzed by Sargent 8c Lundy. The thermal-hydraulic model focused on fan cooler HVS-1C in both units as their higher elevation (with respect to the other three coolers) would cause voiding to begin in these coolers first. Ifno void is formed in HVS-1C, then no void will form in HVS-1A, HVS-1B or HVS-1D or in the associated CCW piping of the respective unit.
Case 1 - Unit 1 MSLB Analysis for CFC HVS-1C Case 2 - Unit 1 LOCA Analysis for CFC HVS-1C Case 3 - Unit 2 MSLB Analysis for CFC HVS-1C Case 4 - Unit 2 LOCA Analysis for CFC HVS-1C
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-18 Attachment Page 7 Two-Phase Flow And CCW Pump NPSH Evaluations The two-phase flowand CCW pump NPSH evaluations reviewd the recovery phase from the case in the previous waterhammer evaluaton that resulted in the highest CCW temperature.
The review determined whether there is a potential for two-phase flow in the CCW system between the CFCs and the main return header.
As the increase in CCW temperature also affects the CCW pump NSPH, a review was made of the transient NPSH margin at the suction to the CCW pumps.
The review of the recovery phase from the worst case transient bounds the recovery from the other three transient cases as well as the quasi-steady state DBA case without LOOP.
The length of the recovery period for the fan cooler unit to reach the quasi-steady state condition is reviewed to determine the timeframe of the recovery.
This review is based on determining the amount of heat stored in the CFC metal tubes and fins by their elevation above normal quasi-steady state temperatures.
The length of the recovery time is shown to be short by comparison of the magnitude of the stored heat with the total CFC heat flux.
The potential for two-phase flow is evaluated at two critical locations in the discharge piping from the CFC; at the highest location in the discharge piping for the HVS-1C cooler and at a position just downstream of the CFC return line throttling valve and flow balancing orifice in the CFC return
- piping, Other postulated locations have increased pressures relative to these two locations; their greater elevation head willexceed the hydraulic pressure drop within the piping to reach the location.
The potential for two-phase flow at each of these locations is determined by evaluating the system pressure based on a hydraulic calculation and determining the saturation temperature for the condition.
This saturation temperature is then compared to the maximum containment temperature to determine whether steam generation willoccur.
Review of the CCW pump NPSH margin is based on setting the NPSHA equal to 1.25 times the NPSHR for the pump (i.e., establishing a calculational margin) and determining the maximum temperature that can be accommodated within the NPSH calculation.
The calculation considered the surge tank elevation head, atmospheric pressure above the vented surge tank, the velocity head in the CCW return header at the surge line tie-in, and the friction losses between the surge tank-tie-in and the pump suction.
The saturation vapor pressure is determined from the classical NPSH formulation and is converted to an equivalent saturation temperature.
This maximum allowable saturation temperature is then compared to the mixed flow condition of the CCW return flows and temperatures from in-service equipment.
The mixing calculation uses the maximum and minimum component flows (as appropriate) from the system operating limits (i.e., CCW system flow balancing requirements) and the maximum calculated water temperature from the CFCs (conservatively based on the containment temperature at the time of pump restart).
Overpressurization of Isolated Piping Sections The evaluation of isolated piping for potential affects of the thermal expansion of trapped fluids determined whether isolated containment penetrations
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-18 Attachment Page 8 and sections of piping internal to containment were subject to overpressurization.
As discussed in Generic Letter 96-06, the potential for systems to fail to perform their safety functions as a result of thermal overpressurization is dependent on many factors.
These factors include leak tightness of valve seats, bonnets, packing glands and flange gaskets; piping and component material properties, location and geometry; ambient and post-accident temperature response; pipe fracture mechanisms; heat transfer mechanisms; relief valves and their settings; and system isolation logic and setpoints. Isolated piping sections potentially affected by thermal overpressurization are identified by reviewing the approximately 160 containment mechanical penetrations as listed in the St. Lucie Unit 1 updated final
'afety analysis report (UFSAR) Table 6.2-16 and the St. Lucie Unit 2 UFSAR Table 6.2.52 and Table 6.2.53.
All piping runs depicted on Piping and Instrumentation Diagrams (P&IDs) were also reviewed.
The fluid media, operating temperature, penetration/pipe section valving, relief valves and connected piping were reviewed to determine the potential for thermal overpressurization during accident conditions.
A number of containment penetrations and piping sections were identified in each unit that are vulnerable to thermal overpressurization as a consequence of design basis accidents (DBA).
Calculations were performed for each of the identified pipe sections.
The calculations for the piping evalations assumed an initial temperature and pressure and a final temperature of 250F based on the long term (2 minutes+) post-DBA containment curves.
ASME Section III,Appendix F is utilized as the basis for acceptability of the affected piping. As stated in section F-1200 of Appendix F, the intent of the evaluation is to demonstrate that there willbe no violation of the pressure retaining boundary. This is accomplished by performing a quasi-inelastic analysis to demonstrate that the hoop strains resulting from the increased internal pressure combined with the bending strains will not result in stress levels that would cause a violation of the pressure retaining boundary.
ASME Section III, Appendix F is utilized as the basis for acceptability of the affected safety-related valves.
This is accomplished by performing a review of valve structural integrity using design inputs for wall thickness, internal diameter, and material. The evaluation is based on review of the maximum hoop stress developed within the valve body as a result of the fluid pressure required to reach yield stress within the associated pipe section.
Within this analysis, the hoop stress is compared to the acceptance criteria delineated in ASME Section IIIAppendix F Section F-1331.1(a),
4.0 Evaluation Waterhammer Evaluation Four cases were analyzed using the thermal-hydraulic model to determine ifvoiding would occur in the CFCs during DBA transients with LOOP. The model focused on fan cooler HVS-1C in both units as their higher elevations would cause voiding to begin first. Ifno void
t St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-18 Attachment Page 9 is formed in HVS-1C, then no void would form in HVS-1A, HVS-1B, HVS-1D, or in the associated CCW piping of the respective unit.
Case CFC Tem CCW Pum Restart-Time frame for Steam Generation Unit 1 MSLB Analysis Unit 1 LOCA Analysis Unit 2 MSLB Analysis Unit 2 LOCA Anal sis CFC water remains below 152F for first 16 seconds CFC water remains below 215F for, first 16 seconds CFC water remains below 169F for first 16 seconds CFC water remains below 187F for first 16 seconds No steam generation occurs in CFC HVS-1C out to 30 seconds No steam generation occurs in CFC HVS-1C out to 16.1 second No steam generation occurs in CFC HVS-1C out to 20.7 second No steam generation occurs in CFC HVS-1C out to 19.3 second The calculation cases were developed to provide an accurate prediction of CCW system response to the containment analysis response curves.
The CFC analysis model formulaton does not provide large analysis conservatisms.
This approach was adopted as the containment pressure and temperature response curves contain conservatisms to maximize the response for an assumed instantaneous pipe rupture of the largest reactor coolant system (RCS) line or main steam line.
Given the non-mechanistic nature of the break scenario and the conservatisms within the containment model, including a large amount of conservatism in the CFC analysis model was not warranted.
I Allowingfor 10% of the manufacturer's design fouling assumed in the DBA LOCA performance data and a normal CCW surge tank level (1 foot higher) lengthens the time to bulk boiling in the Unit 1 LOCA case to 16.6 seconds.
Allowing for the entire amount of fouling used in the vendor's DBA LOCA performance data would lengthen the time to bulk boiling to 17.6+ seconds.
FPL has concluded that no steam generation willoccur in the containment fan coolers for either unit prior to the time the CCW pumps restart (within 16 seconds after LOOP).
Given the inherent conservatisms in the containment response
- curves, these results are acceptable and no steam generation is expected for the scenarios presented within Generic Letter 96-06.
Two-Phase Flow And CCWPumpNPSHEvaluation Of the cases reviewed in the previous section, the Unit 1 LOCA scenario (Case 2) results in the greatest loss of subcooling margin within the CFC fan cooler.
Recovery from this case is of interest with respect to two-phase flow issues.
CCW pump restart results in CCW system pressurization and development of design flow within approximately 1 second.
The repressurization of the CCW system:
a) raises the CCW system saturation pressure at the CFC (increases voiding margin); b) induces flow increasing CFC heat transfer (decreases voiding margin); c) induces return line hydraulic pressure drops (decreases the
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-18 Attachment Page 10 voiding margin); d) induces flow increasing return line heat transfer (increases, then decreases, voiding margin).
The recovery period for the fan cooler unit to reach quasi-steady state is determined to be short (3+
second timeframe); somewhat longer than the 1 second timeframe for full recovery of CCW flow.
The potential for two-phase flow is evaluated at two critical locations in the discharge piping from the CFC; at the highest location in the discharge piping of the HVS-1C cooler and at a position just downstream of the throttle valve and flow balancing orifice in the CFC return piping. The potential for two-phase flow at these locations is determined by evaluating the system pressure based on a hydraulic calculation and determining the saturation temperature for the condition.
This saturation temperature is then compared to the maximum calculated CCW water temperature (conservatively assumed to be the containment temperature at the time the pump starts) to determine whether steam generation willoccur.
This approach accounts for the additional heat exchange of the water during its transit through the CFC cooler and the return piping.
The saturation temperature for the location just downstream of the FE-14-2C flow orifice (262.2F) was shown to exceed the highest containment saturation temperature at the time the pump restarts (260F) and thus two-phase flow does not occur at this location.
The saturation temperature for the location at the highest location downstream of the HVS-1C CFC was shown to be 294.4F which is greater than the maximum containment saturation temperature. Thus two-phase flow willnot occur at this location.
As other postulated locations have increased pressures relative to these two locations, two-phase flow will not occur in the CCW system during the recovery phase from the Unit 1 LOCA case.
Comparison of the Unit 1 LOCA Analysis (Case 2) with the three other DBA cases and the HVS-1C CFC piping geometry with other coolers indicates that the above review bounds the recovery period for all other coolers in both units and for the three other cases, as well as the quasi-steady state design case.
Review of the CCW pump NPSH margin is based on setting the NPSH~ equal to 1.25 times the NPSH~ for the pump and determining the maximum temperature that can be accommodated within the NPSH calculation.
The evaluation showed that an entering CCW pump temperature of 212.3F provides NPSHequal to 125% of NSPlg Accounting for the maximum CFC return line temperature, and maximum or mininimum component flow rates (as appropriate), the maximum mixed temperature at the CCW pump suction is 159.8F.
As the actual temperaure is less than the allowed temperature, the CCW pumps willnot cavitate due to the flushing of hot stagnant CFC water into the CCW pump suction line during the recovery period, FPL concluded that two-phase flow will not occur in the CCW system for the CFC coolers during quasi-steady state conditions or during the recovery phase following the DBA/LOOP cases reviewed
I'
St. Lucie Units 1 and 2 Docket Nos, 50-335 and 50-389 L-97-18 Attachment Page 11 in the previous section.
As a consequence, the hydraulic balance of the CCW system will not be affected and waterhammer is not anticipated.
Sufficient NPSH exists for CCW pump operation during the recovery phase from the worst case transient event.
Overpressurization of Isolated Piping Evaluation Isolated piping sections potentially affected by thermal overpressurization were identified by reviewing all piping systems penetrating containment and those contained within containment.
This review included all containment mechanical penetrations based on a UFSAR list of penetrations and reviews of all piping shown on P&IDs.
The review of P&IDs was intended to identify: a) containment penetrations that have isolation valves both inside and outside containment which are closed or automatically close on safety injection actuation system (SIAS) and/or containment isolation actuation system (CIAS) signals; and, b) fluid filled, double isolated pipe sections within containment, Penetrations and piping sections were reviewed to determine those that would be fluid filled and subject to thermal pressurization.
Screening criteria included systems wholly outside containment; systems penetrating containment or within containment not handling liquids; sections of fluid filled piping inside containment normally operating at higher than post-DBA containment temperatures; systems with thermal relief provided by relief devices, check valves, or solenoid/air operated valves (AOV) with pressure under the seat; sections of piping open to vessels containing compressible fluids or provided with pressure relief devices; etc.
While Generic Letter 96-06 addresses the potential for safety related systems to fail to perform safety related functions due to thermal pressurization, the screening process also identified isolated pipe sections in non-safety related systems not penetrating containment.
Following the screening process a number of penetrations and piping sections in both units were identified as vulnerable to thermal pressurization.
A listing of the identified cases, the piping sections and services involved, and their circumstances is provided in Table 1 for Unit 1 and Table 2 for Unit 2.
Review of the affected penetrations and pipe sections concluded that:
1)
None of the identified penetrations serve a safety-related system whose function is required post-LOCA or MSLB. The safety related criteria is limited to a passive, containment boundary function.
2)
The identified isolated pipe sections within containment are largely for non-safety related systems not required post-DBA, Double isolated, fluid filled pipe sections with safety related functions were identified in the CVCS system drain lines and in the RCS refueling level transmitter line.
4 St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 I 97-18 Attachment Page 12 The identified penetrations and piping sections were reviewed for acceptability under ASME Section III, Appendix F criteria.
Calculations were performed to determine the expected pressure increase within piping sections and the results were reviewed to deetermine the piping condition.
FPL concluded that no violation of pressure retaining boundary willoccur and the identified penetrations and piping sections meet functionality criteria under ASME Section III, Appendix F.
System valves with a safety function were reviewed for the expected pressure increases in isolated piping sections. FPL concluded that all valve models meet functionality criteria under ASME Section III, Appendix F.
During the review of the valves serving Unit 2 penetration 23 and penetration 24 (which provide supply and return CCW piping for the RCP seals, the RCP lube oil coolers, and the control rod drive coolers) the thermal pressurization assumed in the piping evaluation resulted in excessive shear stresses in the stems of the 8" Pratt N-MK-IIbutterfly valves.
While the calculated material stress is less than the yield stress of the stem, it is approximately 14% above the Appendix F acceptance criteria.
Penetrations 23 and 24 were reviewed and determined to be functional as the calculated fluid pressures based on pipe material yield stress willnot occur in practice.
This judgement is based on the reasonable expectation that leakage rates from these resilient seated butterfly valves (i.e., rubber lined) willbe greatly in excess of approximately 0.05 gpm (the leakage required to relieve pressure sufficiently to meet the ASME Section IIIAppendix F acceptance criteria). The resilent rubber lining will result in leakage rates well in excess of this required flow at differential pressures in the range of 500 psid. As the required leakage to limitpressure is a very small fraction of the allowable ILRT leakage criteria (10 gpm at 42 psid) and the valve is expected to reseal as the pressure differential is reduced to design values, containment integrity willbe provided by the penetrations valves.
Consideration of SingleFailures The affect of potential single failures on the evaluation methodologies was considered and the evaluation results are discussed in the following paragraphs.
Voiding/Two-Phase Flow Evaluations - The CCW system within both units is automatically split into two separate and independent safety-related trains by redundant valves which close on safety injection acutation system (SIAS) signal.
Accordingly, the CCW system is inherently resistant to single mechanical failures that would affect both trains.
Redundant electrical equipment (EDGs, motor control centers (MCC), electrical busses, and control circuits) is provided to separately support the
'A'rain and the 'B'rain ensuring that postulated single electrical failures would not affect both safety-related trains.
As the CFC fan is not assumed to coastdown during the timeframe of the analyzed transient, failure of the CFC fan to be stripped from an energized electrical bus has no affect on the evaluation.
Failure of a shutdown heat exchanger outlet valve to open would adversely affect the CCW pump NPSHcalculation but not the NPSHof the CCW pump in the opposite train. As discussed in the plant UFSARs, limitingsingle failures in the electrical system, CFC/CCW system and containment spray system were considered in the development of the containment temperature and pressure response curves used within the evaluation.
These limiting failures would affect the
St. Lucie Units 1 and,2 Docket Nos. 50-335 and 50-389 L-97-18 Attachment Page 13 containment response curves beyond the timeframe of this evaluation.
Based on the above, postulated single failures willnot affect the voiding or two-phase flow evaluations.
Thermal Overpressurization - Thermal overpressurization occurs due to the normally closed position of manual valves or the closure of automatic valves on receipt of an SIAS or a containment isolation actuation system (CIAS) signal. Failure of automatic containment isolation valves to close have been addressed in the design of the containment boundary.
Such single failures would mitigate or preclude the thermal overpressurization cases postulated within this evaluation.
Single active failures of valves that fail to open in response to SIAS or CIAS signals could potentially increase the number of penetrations susceptible to thermal pressurization.
Cases in this category would be limited to valves in safety systems opening to provide flow into or out of containment.
Review of containment penetrations for safety systems with active opening valves (e.g., auxiliary feedwater, containment spray, safety injection) did not identify any additional penetrations that would be subject to thermal pressurization.
In each of these cases, the piping within containment is open ended or provided with a check valve.
Based on the
- above, postulated single failures will not affect the thermal overpressurization evaluation.
5.0 Operability Assessment Review Waterhammer Containment air cooler cooling water systems for St. Lucie Units 1 and 2 are not susceptible to steam generation during loss ofoffsite power coincident with design basis accidents, As steam generation is not expected, waterhammer loads as postulated in GL 96-06 willnot occur.
Two-Phase Flow in Safety-Related Piping and Components Containment air cooler cooling water systems for St. Lucie Units 1 and 2 are not susceptible to two phase flow conditions during quasi-steady state DBA events or following DBA events with a coincident loss of offsite power.
Adequate NPSH is available for CCW pump operation during analyzed transients following DBA heating of containment coolers with LOOP.
OverpressurizationofIsolatedPiping Sections Two (2) penetrations in Unit 1
and six (6) penetrations in Unit2 are susceptible to thermal overpressurization.
Additionally, a number of piping sections within the two containments are vulnerable to thermal pressurization.
An operability review of each containment penetration and those double isolated lines within containment which serve a safety function during power operation (e.g., small bore lines connected to the RCS or CVCS boundaries) was performed.
FPL reviews of the piping sections concluded that each affected line remains operable under the acceptance criteria of ASME Section III, Appendix F with no violation of the pressure retaining boundary.
V St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-18 Attachment Page 14 FPL reviews of the valves serving Unit 2 penetrations 23 and 24 (w'hich provide supply and return CCW for the RCP seal/lube oil coolers and control rod drive coolers) is based on the reasonable expectation that leakage rates from these resilient seated (i.e., rubber lined) butterfly valves willbe greatly in excess of approximately 0.05 gpm.
0.05 gpm is the leakage required to relieve pressure sufficiently to meet the ASME Section III, Appendix F acceptance criteria.
Also, as the required leakage to limitthe pressure is a very small fraction of the allowable 10 CFR 50 Appendix J leakage criteria (10 gpm) for the valves and the valves are expected to remain undamaged and reseal as the pressure differential is reduced to design values, containment integrity willnot be compromised. FPL reviews of the remaining valves concluded that each affected valve remains operable under the acceptance criteria of ASME Section III, Appendix F with no violation of the pressure retaining boundary.
In conclusion, FPL reviews of the piping sections and valves determined that each of the affected lines remain operable under the acceptance criteria of ASME Section III, Appendix F with no violation of the pressure retaining boundary.
FPL considers these operability assessments to be valid for the remainder of the current operating cycle of each unit and ifnecessary, the next operating cycle of each unit.
6.0 Results Summary Containment air cooler cooling water systems for St. Lucie Units 1 and 2 are not susceptible to either waterhammer or two phase flow conditions during postulated accident conditions.
Loss of offsite power coincident with design basis accidents can be accommodated without the initiation of voiding within the cooling water system based on the existing emergency diesel loading sequence.
Two phase fiowconditions do not occur during the transient recovery or steady state design basis conditions.
As a consequence, waterhammer is not anticipated within this system.
Adequate CCW pump NPSH is provided during the recovery trainsient from the worst case event.
Of the 160 containment mechanical penetrations within Units 1 and 2, two (2) penetrations in Unit 1 and six (6) penetrations in Unit2 were identified as vulnerable to potential thermal overpressurization.
Additionally, a,number of isolated pipe sections within the containments are vulnerable to thermal pressurization.
An operability review was completed using the acceptance criteria found in ASME Section IIIAppendix F in accordance with the guidance of Generic Letter 91-18.
All the affected piping and penetrations will remain functional as the pressure boundaries willnot be violated.
As the St. Lucie Units 1 and 2 UFSARs do not recognize the use of Appendix F for design bases conditions, corrective actions are required in the longterm.
Preliminary review indicates that adoption of ASME B&PV Code,Section III, Appendix F via a license ammendment may be acceptable as long-term design bases for Service Level D evaluations.
This is the planned corrective action.
h
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-18 Attachment Page 15 Barring this approach, corrective actions to change the field situation are required to address thermal pressurization of the isolated piping sections.
Implementation of alternative corrective actions will require sufficient lead time for design development, budgeting, ordering of ASME materials and relief valves, and outage planning.
7.0 Corrective Actions FPL considers the operability assessments discussed in Section 5.0 of this report to be valid for the remaider of the current operating cycle of each unit and ifnecessary, due to NRC review schedules, the next operating cycle of each unit.
2.
FPL willsubmit license amendment requests to incorporate ASME BEcPV Code,Section III, Appendix F, as long-term design basis for Service Level D evaluations into the licensing bases of St. Lucie Units 1 and 2.
The evaluations necessary for this applicaton will require the detailed reanalysis of each of the affected penetrations or piping sections.
The engineering analysis is expected to take six to eight weeks, plus an additional 4 weeks for the Technical Specification required safety reviews.
Consequently, FPL plans this submittal April 30, 1997.
3.
Should FPL, during the preparation of the proposed license amendments, identify any of the piping configurations subject to thermal overpressurization that will not meet the planned criteria for use of ASME Section III Appendix F, FPL will implement modificatons to corrective the field conditions.
A.
For St. Lucie Unit 1, any such changes will be implemented during the upcoming Fall 1997 refueling/steam generator replacement outage (SL1-15).
B.
For St. Lucie Unit 2, the next refueling outage (SL2-10) is scheduled to start on April 14, 1997.
Since the detailed reanalysis of the affected piping sections may not be completed until the end of March 1997, there is insufficient time to identify the need, complete the detailed engineering design, procure materials, and plan for the installation of any corrective actions during the upcoming Unit 2 outage.
FPL proposes to delay any such corrective actions on Unit 2 until at least the next refueling outage (SL2-11), which is currently scheduled for the Fall 1998.
4.
Should NRC review of the proposed license amendment discussed in corrective action 2 not be approved, FPL willprovide a plan and schedule for implementing alternative corrective actions within 90 days of the date of the NRC disapproval.
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 I 97-18 Attachment Page 16 Table 1 - St. Lucie Unit 1 Penetrations and Piping Vulnerable to Thermal Overpressurization Containment Penetrations Vulnerable to Thermal Overpressurization Pcn //
Section Sch Material Class Circumstances 47 3&S-52 1&S-82 3ZS-56 1&S-83 40 80 40 80 304SS 304SS 304SS 304SS Refueling Pool Purification Supply Line Rcfucling Pool Purification Return Linc 2 NC manual gate valves, onc IC and one OC with test connection 2 NC manual gate valves, onc IC and onc OC with test connection Piping Sections Inside Containment Vulnerable to Thermal Overpressurization Near Section Sch Material Class Service Circumstances PRZ 3/4-RC-131 304SS A
Pressurizer Vent PS isolated by NC manual globe valves RCS CHG 42 3/4-RC-144 3/4-RC-376 3/4-RC-378 3/44"H-180 3/~H-B07 3/4TH-B08 3/44 H-B09 3/4TH-B17 1&H-B26 14H-B27 1ZH-B28 3&&45 3&ST 2-C&42 1 "vents 160 160 160 160 160 160 160 160 160 160 160 40 10 40 304SS 304SS 304SS 304SS 304SS 304SS 304SS 304SS 304SS 304SS 304SS 304SS 304SS 304SS A
A B
B D
D RCS Lcvcl Transmitter Various CVCS Vents &Drains Reactor Cavity Sump Pumps PS isolated by manual globe valves, section contains orifice with A [ B class brcak In each case, PS is isolated by double manual globe valves.
As single manual valve is required for B ( D class brcak, PS can be dcclassiiicd or upstream valve opened.
PS isolated by 2 NC AOVs (globe valves) OC, NNS closed piping IC is thinner wall; willprotect pcnctration 47 3-CS-55 10 304SS D
Refueling Pool Purification Return PS isolated by 2 NC manual gate valves NC - Normally Closed, SV - Solenoid Valve, AOV-AirOpcratcd Valve, MV-Motor Valve, IC-Inside Containmcnt, OC - Outside Containmcnt PS - Pi Section
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-18 Attachment Page 17 Table 2 - St. Lucie Unit 2 Penetrations and Piping Vulnerable to Thermal Overpressurization Containment Penetrations Vulnerable to Thermal Overpressurization Pen 4 Section Sch Material Class Service Circumstances 23 8&C-138 vent linc 8-CC-168 vent line 40 CS 40 CS CCW Non-Essential Header Supply Linc CCW Non-Essential Header Return Linc AOV OC &AOVIC close on SIAS AOVOC &AOVIC close on SIAS 42 3-CRIES 2&S-95 1-CS-93 40 40 80 304SS 304SS 304SS 3/44.H-129 160 304SS Reactor Cavity Sump Pump Discharge to Equipment Drain Tank RCP Controlled Bleed-Off Linc to VCT NC AOV OC & NC AOVIC close on SIAS/CIS AOV OC &AOVIC close on CIS 46 47 3-CS-52 1-CS-82 3&S-56 1-CS-83 40 80 40 80 304SS 304SS Refueling Pool Purification Supply Linc Rcfucling Pool Purification Return Line 2 NC manual gate valves, onc IC and onc OC with test connection 2 NC manual gate valves, onc IC and one OC with test connection Piping Sections Inside Containment Vulnerable to Thermal Overpressurization Near Section Sch Material Class Service Circumstances PRZ 3/4-RC-861 160 304SS PRZ Vent PS isolated by NC manual globe valves RCS 42 3/4-RC-376 3/4-RC-378 2&S-94 3<844 3&S-106 ctc 2-PMW-7 2-PMW-10 3-CS-53 ventline 160 160 40 10 10 80 40 40 40 40 10 304SS 304SS 304SS 304SS 304SS 304SS 304SS 304SS 304SS 304SS 304SS A
B D
D D
B D
~
D D
D D
RCS Level Transmitter Reactor Cavity Sump Pumps Primary Water Make-Up Rcfucling Pool PuriTication Supply PS isolated by manual globe valves, section contains oriTiec with A ) B class brcak PS isolated by NC AOVIC and pump check valve PS isolated by AOV OC and check valve IC. NNS closed piping IC is thinner wall; willprotect pcnctration PS isolated by 2 NC manual gate valves 46 3&S-53 10 304SS C
tD)
Rcfucling Pool Purification Supply PS isolated by 2 NC manual gate valves, Class D for power operation 47 3&S-109 vcntline 10 304SS D
Rcfucling Pool Purification Return PS isolated by 2 NC manual gate valves 47 47 3&S-55 40 3&S-54 40 304SS 304SS C
P)
Rcfucling Pool Purification Return Refueling Pool Purification Return PS isolated by 2 NC manual gate valves PS isolated by 2 NC manual gate valves, Class D for power operation NC - Normally Closed, SV-Solenoid Valve, AOV-AirOpcratcd Valve, MV-Motor Valve, IC-Inside Containmcnt, OC - Outside Containmcnt PS - Pi Section