ML17309A612
| ML17309A612 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 01/30/1997 |
| From: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| To: | Vissing G NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-96-06, GL-96-6, NUDOCS 9702070236 | |
| Download: ML17309A612 (64) | |
Text
CATEGORY j.
L REGULATOlOINFORMRTION DISTRIBUTION TEM (RIDE)
ACCESSION HBR:9702070236 DOC.DATE: 97/01/30 NOTARIZED: YES DOCKET I FACjL:50-)44 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas & Electric Corp.
RECIP.NAPE RECIPIENT AFFILIATION VISSINGFG;S. Document Control Branch (Document Control Desk)
SUBJECT:
Responds to GL 96-06, "Assurance of Equipment Operability &
Containment Integrity During DBA Conditions."
DISTRIBUTION CODE: A072D COPIES RECEIVED:LTR ENCL SIZE: A TITLE: GL 96-06, "Assurance of Equip Oprblty a Contain.Integ. during Design T
NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 E
RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL NRR/WETZEL,B. 1 1 PD1-1 PD 1 1 VIRGINS G. 1 1 INTERNAL: ILE CgHTZR 0 1 1 NRR/DE/EMEB 1 1 NRR/DSSA/SCSB 1 1 NRR/DSSA/SPLB 1 1 Y EXTERNAL: NOAC 1 1 NRC PDR 1 1 D
U NOTE TO ALL MRIDS<< RECIPIENTS PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK/
ROOM OWFN 5D-5(EXT. 415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
TOTAL NUMBER OF COPIES REQUIRED: LTTR 9 ENCL '9
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ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N.K 14649 0001 TCLCPHONC ARcA coDc %la 546-2700 January 30, 1997 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Guy S. Vissing Project Directorate I-1 Washington, D.C. 20555
Subject:
Response to Generic Letter 96-06, "Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," dated September 30, 1996 R.E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Vissing:
Rochester Gas and Electric (RG&E) Corporation has completed its evaluation of the R.E.
Ginna Nuclear Power Plant as requested by Generic Letter 96-06. The purpose of this letter is provide the subject evaluation within the requested 120 day response (the generic letter was received by RG&E on October 2, 1996) ~
The attached evaluation addresses the effects of potential water-hammer and two-phase flow with respect to the service water supply to the containment recirculation fan coolers (CRFCs).
Also included is an assessment of the potential for thermally induced overpressurization of isolated piping sections. The results of these two evaluations are presented in the attached report (Sections 2.1 and 2.2 for the CRFCs and Section 2.3 for isolated piping sections).
Very truly yours,
'tt702070236 970i30 Robert C. Mecre y PDR ADOCK 05000244 Vice President, uclear Operations P PDR Subscribed and sworn/affirmed before me this 30th day of January, 1997 Nota ublic SHARON J. ECKERT Notary Public in the State of New Yorft My Commission Exptres March 29, Monroe County
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xc: Mr. Guy S. Vissing Project Directorate I-1 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King oE Prussia, PA 19406 Ginna Senior Resident Inspector
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ATTACHMENT 4.6 GL 96-86 EVALUATION OF PENETRATION PIPING SYSTEMS FOR THERMALLY INDUCED OVERPRESSUF TION CON PENETRATION AFTER ACCIDENT SIZE P&ID DWG.:33813- COMMENTS PENETRATION S
SIZE P &ID DWG.:33813- COMMENTS AFTER ACCIDENT NY RN HUT WN. MPTY TH R S 286o 3 8 1278-1 RU PEP N A 18 4-2 286b 1277-1 NO PM NT P T M N 1 4 MPTY URN P R ON 287o 3 8 278-1 REU NEEDE N NTRAPM NT A T A A R F 28 b 277-1 NO I PM NT OPENT S MGN
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46-1 A 1882 T fPR UR A N RAP M T P F R 316 1258-14-1 A ~
NQ IA PMFNT OP N F R SW F W U LY N A 5HUT 125 1246-1 NTRAPM NT N F R W W T EO 1 4 1 P
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1 128 46-1 321 12 -1 M N T. 0 NT T M N 1
322 127 -1 NQ MENT OP N T STEAM N RATOR 24 148 1247. N N TRAP M N ' N T RHR P P N A A H 1 8-1 4 1 T M PAPM NT N T 14 1 7 1 P T NO 281o 1 8- 14 NQ lb 8- M N N T PNT M N 8'884-1 1
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RESPONSE TO GENERIC LETTER 96-06 "Assurance Of Equipment Operability And Containment Integrity During Design-Basis Accident Conditions" II For R. E. Ginna Nuclear Power Plant Rochester Gas & Electric Corporation Rochester, New York 1 4649 97P2Q70236
S I
,7 l TABLE OF CONTENTS Page 1.0 Introduction 1 2.0 Evaluation of GL 96-06 Issues For Ginna Station..... 3 2.1 Susceptibility of Ginna Containment Recirculating Fan Coolers (CRFC) Water System To Water Hammer During Postulated Accident Conditions .... ~ 3 2.1.1 Description of Ginna CRFC's 3 2.1.2 Description of Ginna CRFC's Cooling Water S ystem..... .. ~ ~ . ~ ~ ...... ~ ~
2.1,.3 Evaluation of CRFC's During LOCA With Concurrent Loss Of Offsite Power (LOOP).....
'2.1.3.1 Description Of Event Scenario.........
2.1.3 2 Evaluation Of Structural Integrity
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Of Tubes Due To Water Hammer.. ., ~
2.1.3 3 Evaluation Of Water Hammer Potential
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Considering Effects of Inlet &. Outlet P I ping o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 2.1.3.4 Summary - Water Hammer Evaluation.. 7 2.2 Two-Phase Flow In Safety-Related Piping And C omponents............. ~.......:.~......... .
~ ~ ~ ~ ~ ~ ~ .. ..~.......
~ 8 2.2.1 Design Of Enhanced CRFC's. ~ .. . ..:..~.....~....
~ ~ ~ 8 2.2.2 Evaluation Of CRFC's To Two-Phase Flow F 0 I ITIatI On ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
2.2.3 Evaluation of Service Water System (SWS)
Piping To Two-Phase Flow Formation.~........, 10 2.2.4 Evaluation Of Water Hammer in SWS Discharge Piping System.......................~.... 11 2.2.5 Summary - Two Phase Flow Formation......... 12 2.3 Overpressurization Of Isolated. Piping...... ..~.......... ~ 14 2.3.1 Piping Systems Inside Containment Susceptible To Thermal Overpressurization.... 14 2.3.1 1 Containment Spray (CS) Charcoal
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Filter Dousing Line........~...............
2.3. 2 Containment Penetrations Susceptible To Thermal Overpressurization..........'... . ..~...... ~ ~ 16 I
TABLE OF CONTENTS Page
'2.3.2.1 Penetration 121a - Pressurizer Relief Tank Makeup Water. ~ ~ ~ ~ ~ ~ ~...... 16 2.3 2.2 Penetration 205- Reactor Coolant
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System (RCS) Loop B Hot Leg S ample.. ........... ............. .... ......
~ ~ ~ ~ ~ 17 2.3.2.3 Penetration 206a - Pressurizer L'qutd Sample............................... 18 2.3.2.4 Penetration 207a - Pressurizer Steam Sample...... .. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 19 2.3.2.5 Penetration 307- Fire Service W atero ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 20 2.3.2.6 Penetration 324 - Demineralized WaterI ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 20 2.3.2.7 Summary - Overpressurization Of I P)ping......,........................
Pl
'olated 22 3 0 List Of References..~.......~..............
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 23
.0 List Of Attachments............ .. .. ~ ~ ~ ~ ~..................... ~ ~ ~ ~ ~ ~ .. 26
RESPONSE TO GL 96-06 R. E Ginna
~ Nuclear Power Plant Rochester Gas 5 Electric Corporation 1.0 Introduction Through Generic Letter 96-06 (Ref. 3.1), the U.S. Nuclear Regulatory Commission (NRC) notified holders of operating licenses of nuclear power reactors of three issues of concern:
(1) Cooling water systems serving containment air coolers maybe exposed to the hydrodynamic effects of water hammer during a loss-of-coolant accident (LOCA) or a main steamline break (MSB). These cooling water systems may not have been designed to withstand hydrodynamic effects of water hammer and corrective actions maybe needed to satisfy system design and operability requirements.
(2) Cooling water systems serving the containment air coolers may experience two-phase flow conditions during postulated LOCA and MSLB scenarios. The heat removal assumptions for design-basis accident scenarios were based on single-phase flow conditions.
Corrective actions may be needed to satisfy design and operability requirements.
(3) Thermally induced overpressurization of isolated water-filled piping sections penetrating containment could jeopardize the ability of accident-mitigating systems to perform their safety functions and could also lead to a breach of containment integrity via bypass leakage. Corrective actions maybe needed to satisfy system operability requirements.
The generic letter also explained specific scenarios and events that have been reported and evaluated to have generic implications that formed the basis of the three issues of concern liqted above.
lg GL96-06 Response
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~ ~ Ginna 2 'age RG&E immediately formed a team to review specific configurations, piping drawings, operability data, design basis commitments and documentation, component and system characteristics, PAID's, and other relevant information for its Girina Station as related to the three issues of concern. Reported here are the findings and actions taken by RGS.E to assure operability of affected systems and components as, requested by GL 96-06 (Ref. 3.1).
GL 96-06 Response R. E. Ginna Page 3 2.0 Evaluation Of GL 96-06 Issues For Ginna Station 2.1 Susceptibility Of Ginna Containment Recirculating Fan Coolers (CRFC's) Water System To Water Hammer During Postulated Accident Conditions Ginna is currently using enhanced design CRFC's that provide cooling of the containment under both normal operating and design-basis accident conditions. One of the design enhancements'for the new coolers is the replacement of the original copper tubes with AL-6XN stainless steel tubes that are stronger and more resistant to erosion/c'orrosion .
2.1.1 Description Of Ginna CRFC's The original CRFC's were replaced by RG5E during the 1993 refueling outage (Ref. 3.2), due to problems such as performance limitations, degrading material condition, and poor accessibility for inspection, cleaning and repair. The replacement CRFC's enhanced the structural integrity and cooling capacity of the coils and also involved substantial modifications to the support structure.
Description of 'these enhancements are documented in a design report (Ref. 3.3) from which the design parameters for the coils that are shown below were taken.
Tube Material ... ~ ~... ..
~ ~ ~ ~ ~ ~ .. ..
~ ~ . AL-6XN (21% Cr, 26% Ni Austenitic Stainless Steel)
No. of Tube Rows (Depth)....~..... ~ . .12 No.'f Tube. Rows (Height)....... .. ~ ~ . 20 Tube Outside Diameter ~......~... .. ~ ~ . 0.625" Tube Thickness 0.035" F inned Length 120 "
Water Flow Circuit ..... ..~.....~..........
~ Triple Serpentine
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Page 4 Fin Material ~... Copper Fin Spacing 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8/inch F'n Thickness .... .. .~...~ ~ ~ . ~ ~.... ~ . ~ ~ .. ~ . ~ ~ 0.010" H eaders . .............. ..
~ ~ ~ ~ ~ .......... ......
~ Stainless Steel Water Box Type There are four CRFC units. System arrangement of these units are, depicted in RGS.E PAID Drawing 33013-1250 (Ref. 3.4).
2.1.2 Description Of Ginna CRFC s Cooling Water System Cooling water for the "A", "B", "C", and "D" CRFC's is supplied by the service water system which utilizes four (4) pumps (Ref. 3.4).
Each CRFC is supplied through a separate penetration assembly with its own service water cooling circuit. The heated cooling water is then piped out of another containment'enetration where it passes through a discharge orifice and joins the others in the Intermediate Building to a 14" header. This header eventually empties into the discharge tunnel (Ref. 3.5) which is open to the atmosphere.
The inlet and outlet cooling water piping of each CRFC are shown in Attachments 4.1 to 4.4. A typical arrangement of this piping with respect to the penetrations, supply and discharge headers is shown in Attachment 4.5.
It is noted that unlike some other plant designs, the Ginna inlet and outlet piping to each CRFC forms a U shape at the penetrations due to the elevational differences (see sketch in Attachments 4.1 to 4.4).
Considering elevations of this piping with respect to the location of the CRFC's, these loops assure that cooling water fills the inlet and outlet piping inside containment when the service water pumps stop for any reason.
I GL 96-06 Response R. E. Ginna
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Page 5 2.1.3 Evaluation Of CRFC's During LOCA With Concurrent Loss of Of Offsite Power (LOOP)
This section evaluates the thermal and dynamic events during a scenario that is postulated in GL 96-06 (Ref. 3.1) which challenges the structural integrity of the CRFC's. The assessment considers specific configuration of the service'water cooling system and
'lant guidelines of the NRC publication, NUREG/CR-5220 (Ref. 3..6) and EPRI reports (References 3.7 and 3.8).
2.'1.3.1 Description Of Event Scenario It is assumed that the CRFC's are functioning normally in steady state condition concurrent with a design-basis accident condition (LOCA or MSLB) when a sudden loss of offsite power (LOOP) occurs.
With a LOOP, the service water pumps on the cooling water side and the fans on the air side of CRFC's will experience a power loss and start to coast down. Since the pumps have little inertia, they will stop in a few seconds. The CRFC fans on the other hand, will coast down for a period of approximately 30 seconds (Ref. 3.1) until the diesel emergency power restarts the service water pumps. During this period, the conditions exist inside the CRFC's that could contribute to boiling due to heating effects of the high temperature
.air/steam mixture in the air side.
Based on computational results of References 3.7 and 3.8, the water inside the coils will indeed boil and generate two-phase steam/water volume inside the tubes. However, with the U bend configuration of both inlet and outlet piping of the service water system, and utilizing guidelines in the NRC publication (Ref. 3.6), it is highly improbable that waterhammer can occur. This is discussed in Section 2.1 3 3. However, the tubes were evaluated assuming that
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waterhammer does occur (see next section).
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Page 6:
2.1.3.2 Evaluation Of Structural Integrity Of CRFC Tubes Due To..
Waterhammer In the unlikely event that steam formed inside the tubes suddenly mixes with cold water supplied by the pumps during restart of the, service water system flow, a deterministic evaluation of the peak pressures and the resulting hoop stress at the tubes was made by RG8cE ( Reference 3.9). The analysis concluded that the condensation-induced water hammer will not adversely affect the structural integrity of the austenitic stainless steel CRFC tubes. It is noted that the CRFC's were changed during the 1993 refueling outage with enhanced structural and heat transfer design capabilities (Ref. 3.2). The CRFC gaskets, cover plates, pass ribs and spacers are held together by multiple bolts that can absorb a substantial amount of strain energy. Consequently, other sections of the CRFC's will also resist the effects of waterhammer peak pressures (Ref. 3.26).
2.1 3.3
~ Evaluation Of Waterhammer Potential Considering Effects Of Inlet And Outlet Piping As shown in Attachments 1 to 5, inlet and outlet piping for each CRFC form a U bend configuration that assures an ample supply of service water between the CRFC tubes and the inlet/outlet headers during the LOOP event scenario. Reference 3.8 has documented the fact that "water on either side of the CRFC's is significantly heated during the boiling phase in the tubes." The mode of heat transfer is predominantly convection as steam inside the tubes diffuses through the inlet and outlet sections.
L The U bend configuration on both sides of the CRFC assures that heated water acts as a "buffer" or transition region between the cold service water from the pumps and the steam voids at the tubes and vicinity. Consequently, the possibility of condensation-induced water hammer is dramatically reduced according to guidelines in Ref. 3 6, ~
'.GL 96-06 Response
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Page 7
~ Ginna since the cold incoming water cannot trap and condense the steam VOIds.
2.1.3.4 Summary - Waterhammer Evaluation During LOOP coincident with design-basis accident conditions (LOCA or MSLB), waterhammer is not a concern for Ginna because'f the following reasons:
(1) In the event that condensation-induced waterhammer does occur inside the CRFC tubes, the enhanced structural capability of the tubes (austenitic stainless steel) will be able to withstand the
'impact of the peak pressures such that the structural integrity of the CRFC will not be affected.
(2) Presence of U bend configuration in the inlet and outlet piping of each CRFC at the penetrations assures a heated "buffer" or transition region preventing cold service water from trapping steam voids that formed inside the tubes and its vicinity. Consequently, condensation-induced waterhammer is avoided.
GL 96-06 Response R. E. Ginna Page 8 2.2 Two-Phase Flow In Safety-Related Piping And Components This section will evaluate Ginna safety-related components, (particularly the CRFC's) and whether these are susceptible to two-phase flow at steady state conditions when operating during design-basis accident conditions. For this particular issue, GL 96-06 concerns are listed below.
(1) Two-phase flow in cooling water systems associated with CRFC's significantly interfere with the ability of the coolers to remove heat under design-basis accident conditions, and can interfere'with the cooling of other safety-related components.
(2) Cooling water systems were designed assuming single-ph'ase flow conditions and containment heat transfer analyses are based on this assumption. Two-phase flow is a much more complex situation to deal with analytically than single-phase flow and involves additional hydrodynamic loading considerations as well as flow, heat transfer, systems interaction and erosion considerations.
, (3) Additionally, the steam that is formed during two-phase flow can accumulate in the cooling water system, restricting flow and potentially resulting in waterharnmer.
2.2.1'esign Of Enhanced CRFC's As was mentioned earlier, the Ginna CRFC's were replaced during
~ the 1993 refueling outage with the design of the replacement CRFC's substantially enhanced, to take care of several problems that were encountered during operation. Among these were'some performance limitations (Ref. 3.3). Concerns for performance limitations when the CRFC's are operating during design-basis accident conditions have been investigated and documented in Reference 3.10, particularly
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Page 9 the following topics that are closely related to GL 96-06 concerns for two-phase flow.
(1) Determine if flashing, steam binding and water hammer can occur in the CRFC coils and SWS discharge piping.
(2) Address the concerns associated with flashing in the CRFC coils and/or SWS discharge piping.
Resolution of these issues, as depicted in Ref. 3.10, were based on conservative thermal performance data and assumptions, which are:
(1) No fouling in CRFC tubes.
(2) 162'/o design margin for heat removal capacity of coils.
(3) SWS inlet water temperature maximum design temperature of 80 F.
The enhanced CRFC design objectives were set such that operation will not result in boiling at the upper bound heat removal rates (Ref.
3.3). Test data for the newly installed CRFC's were then utilized to show margins for avoidance of flashing at the CRFC tubes (Ref.
3.10).
2.2.2 Evaluation Of CRFC's For Two-Phase Flow Formation Based on comparison of test data and results of calculation for CRFC operating requirements,-the following conclusions are documented in an RGS.E Design Analysis (Ref. 3.10):
t (1) Steam binding of the CRFC's will not occur. Voiding will not occur during two pump SWS operation.
GL 96-06 Response R. E Ginna
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Page 10 (2) For a one pump SWS operation, no voiding will occur at CRFC "A", "B", and "D". CRFC "C" could flash by a very small amount, i.e., marginal and within instrument uncertainty, only during- very
,conservative operating conditions (0% fouling, 162% design margin heat rate, 5 80 F inlet SWS water temperature). It is anticipated that this marginal situation will not compromise flow or heat removal . ~
Note that CRFC "C" has already been operating for approximately three years. Consequently, some degree of fouling (i.e., about 1/2%
to 1%) is already present such that it will eliminate the marginal condition of voiding at the outlet of CRFC "C".
(3) If flashing does occur at CRFC "C" discharge, Reference 3.10 has evaluated that this will result in increase of inlet pressure through the lower coils resulting in an offset, i.e., prevention of the flashing event. Even at the reduced flow rate during equilibrium conditions the CRFC's can still meet the minimum heat removal rates assumed in the accident analyses (Ref. 3.11) ~
2.2.3 Evaluation Of Service Water System (SWS) Piping To Two-Phase Flow Formation Evaluation of test data and CRFC design requirements for the conservative conditions that were previously identified has documented the following findings per Reference 3.10.
(1) For a two pump SWS mode of operation, flow and pressure conditions are high enough such that flashing will not occur in any CRFC discharge piping.
(2) During a one pump SWS mode of operation, voiding may occur in the piping downstream of the CRFC's. When this happens, the system will be self regulating. Flashing will reduce the flow through affected CRFC due to increased frictional loss. Consequently,
'he CRFC inlet and discharge pressure will increase supressing void formation. Reduced flow rate has been evaluated in Ref. 3.10 and
GL 96-06 Response R. E. Ginna, Page 11 found that it can still. meet the minimum heat removal rates assumed in the accident analysis.
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(3) Calculations in Reference 3.3 showed that containment temperature will decrease from 286 F to 250 F after the first 2.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> into the design basis accident. The exit temperature from the CRFC's for. the limiting heat removal condition will decrease to less than 212 F when the containment temperature becomes 250 F.
Consequently, two-phase flow will not be a concern after the first 2.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> into the design basis accident.
2.2.4 Evaluation Of Waterhammer in SWS Discharge Piping System The possibility that two-phase flow can occur in the SWS i discharge piping downstream of the CRFC's also creates a potential for waterhammer occurence for certain conditions of operation, i.e.,
one pump SWS mode of operation. This concern has also been evaluated in Reference 3.10. Results are summarized below.
(1) Flashing will most likely occur downstream of the flow orifices for the CRFC's (Ref. 3.5) due to higher elevation and lower system pressures at this location.
(2) Waterhammer can only occur when the two-phase fluid gets trapped by a cooler subcooled fluid (Ref. 3.6). This can potentially happen only at the 14" header where individual 8" CRFC discharge lines join the common line.
(3) Cooler subcooled fluids can come from an inactive coolers, i.e.,
coolers with the fans off, which may happen during certain accident scenarios. Hence, possibility of waterhammer is greatest during this scenario due to limited heat removal, i.e., less service water heatup, by the inactive coolers.
0 GL 96-06 Response R. E. Ginna Page 12 (4) Evaluation of these inactive coolers reveal that with the fan off, they will still remove heat during design-basis accident conditions, where the dominant mode is "condensation heat transfer". This mode is less reliant on forced air flow for heat removal from the air/steam mixture. Since steam binding at the coils will not occur, the SW exit temperature from these "inoperative coolers" will be greater than 80 F, thereby reducing the difference in exit temperatures among the CRFC's. Potential for waterhammer in this condition consequently becomes negligible.
(5) Since the flow orifices are located in the Intermediate Building, piping inside the Containment Building is not susceptible to water hammer for any mode of SWS operation.
I 4 2.2.5 Summary - Two Phase Flow Formation in Safety-Related Piping and Components Evaluation of the susceptibility of the CRFC cooling water system to two-phase flow formation during design-basis accident conditions has revealed the following results:
(1). For two pump SWS mode of operation, two-phase flow formation will not occur anywhere in the CRFC cooling water piping and tubes.
(2) For one pump SWS mode'of operation, CRFC "C" has a marginal and a very low probability tendency to develop voiding at the cooler discharge based on very conservative conditions of operation, i.e.,
0% fouling, maximum heat transfer loads, and 80 F cooling water inlet temperature. However, if this were to occur, flow readjustment is expected to increase back pressure at the CRFC's to protect the tubes from steam binding. The reduced flow rate has been found to still remove design-basis accident heat loads required by the Ginna
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Page 13 UFSAR (Ref. 3.10).
(3) For one pump SWS mode of operation, SWS piping downstream of the discharge orifice has a possibility to flash into steam. Due to increased friction, flow will consequently decrease. Analysis of this condition shows that required heat removal rates are still met. The reduced flow results in an increase of back pressure at the coils which assures that two-phase flow does not develop at the coils or tubes (Ref. 3.10).
(4) There is a slight possibility that waterhammer can occur at the 14" discharge header during a scenario where one or more CRFC's is out of operation (i.e., fan is off). However, the condensation mode of heat exchange at the tubes will still transfer heat to the SWS cooling liquid thereby raising its exit temperature as it arrives at the discharge header. This condition will not only mitigate the effects of water-hammer, but also decreases substantially its probability of occurence.
(5) Waterhammer will not occur in the SWS piping inside the Containment Building.
GL 96-06 Response R. E. Ginna Page 14 2.3 Overpressurization Of Isolated Piping The third issue of concern in GL 96-06 arose out of several reported experiences by utilities and the fundamental concept that heated water while it is trapped is capable of producing extremely high pressure due to its thermal expansion (Ref. 3.1). The high pressure has a potential to fail the pipe and if the pipe is part of accident-mitigating systems, then the safety function of the system is jeopardized. The potential for thermally induced overpressurization is greatest during design-basis accident conditions inside containment since the inside temperature can rise up to 286 F for Ginna over a 10 second period.
Of particular interest and concern for this issue are safety-related systems inside containment such as containment spray, and piping that penetrate the containment. An RG&E team reviewed safety and nonsafety related piping systems inside containment that maybe exposed to this particular sequence of events, including all penetration assemblies (Ref. 3.12). The interest in nonsafety related sys'tems is solely dictated by interaction potential with safety related ones. Results of the evaluation are documented in this section.
2.3 ~ 1 Piping Systems Inside Containment Susceptible To Thermal 0verpressurization P&ID's for systems inside containment were reviewed, particularly those that are required to operate during design-basis accident conditions. The review was extended to nonsafety related systems that may attached to safety related systems. Rupture of these systems may divert flow, thereby affecting the effectiveness of accident-mitigating systems. One system has been found at Ginna that meets this scenario.
GL 96-06 Response R. E Ginna
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Page 15 2.3 1.1
~ Containment Spray (CS) Charcoal Filter Dousing Line Review of P&.ID Drawing 33013-1261 (Ref. 3.13) revealed that during the injection phase of the Containment Spray Piping System of a design-basis accident condition, cold water can get trapped in the CS Charcoal Filter Dousing Line. Subsequent thermal heatup to a maximum containment temperature of 286 F (Ref. 3.14) can generate high pressures that may potentially rupture the piping (Ref.
3 15) and subsequently divert some of the CS flow, reducing the
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amount of CS water delivered in the CS ring headers. It is noted that CS system has the primary function of reducing containment pressure during design-basis accident conditions using cold water sprays to condense the steam inside containment. An Action Report (Ref. 3.18) was written to formally track and assess the problem.
A. Operability Of Affected System Upon preliminary identification of the potential problem, the CS Charcoal Filter Dousing Line was immediately isolated and vented to assure the operability of the CS Piping System, by eliminating the possibility of thermal overpressurization and potential rupture of the charcoal dousing line (Ref. 3.16). Isolation and venting of the Dousing Line was made at about 0017 EST on Dec. 21, 1996. At this time, preliminary evaluations revealed that overpressurization would not results in rupture of the piping. When formal calculations were completed on Dec. 23, 1996 and found that there would be
'potential rupture of this line, the NRC Operation Center was notified (Ref. 3.16).
B. Actions To Prevent Recurrence On January 3, 1997 the CS Charcoal Dousing Line was.modified by installation of a relief valve (Ref 3.17) and the line was returned to its normal alignment. An LER was written (Ref. 3 16) and submitted
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to the NRC on January 22, 1997.
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2.3.2 Containment Penetrations Susceptible To Thermal 0verpressurization The RGS.E team has evaluated all penetration assemblies (Ref.
3.12) considering the possibility that cold water can get trapped during testing, normal operation, or maintenance and subsequently heated during design-basis accident conditions or even during seasonal change and exposed to thermal overpressurization of the assemblies. For completeness, PS.ID's that depict the penetrations and piping systems inside containment were also reviewed. Summary of the evaluation is shown in Attachment 4.6, which depicts six (6) penetrations that needed further investigations. These evaluations are reported below.
2.3:2.1 Penetration 121a - Pressurizer Relief Tank Makeup Water Review of the P&ID drawing (Ref. 3.19) and Figure 6.2-29 of Reference 3.12 revealed that makeup water, which is at relatively cold temperature may be trapped between Valves 548 and 508. A section of trapped fluid can get heated during design-basis accident conditions thereby subjecting piping and valves in the penetration assembly to thermal overpressurization that may lead to potential rupture. of the pipe or damaging the inside CIV (Ref. 3.20).
A, Operability Of Affected Penetration Assembly A more detailed evaluation of the Valve 548 internals showed that before the pipe ruptures, the thermally generated fluid pressure can
0 GL 96-06 Response R. E. Ginna Page 17 overcome the opening force required to lift the plug. This assures that thermal overpressurization will not degrade the mitigating capability of the penetration assembly because of the relief action of Valve 548.
B. Action Planned For Permanent Protection Of Penetration Assembly A Plant Change Request (PCR) is currently being planned to install a relief valve on the PRT Makeup System to permanently protect the penetration assembly 'from thermal overpressurization.
2.3.2.2 Penetration 205 - Reactor Coolant System (RCS) Loop B Hot Leg Sample Review of P&ID (Ref. 3.21) and Figure 6.2-52 of Reference 3.12 showed that hot leg sample can get trapped in the penetration assembly. An engineering analysis (Ref. 3.20) found that assuming
. Valve 955 (Ref. 3.21) does not leak, the trapped fluid will be at a higher temperature than the maximum containment temperature of 286 F. Consequently, fluid contraction from RCS temperature to containment ambient temperature will be greater than fluid expansion from ambient to 286 F so that no thermal overpressurization results.
However, the RGS.E team raised the issue of a postulated small leak at Valve 955 which can fill up the line and may subject the assembly to thermal overpressurization. Consequently,. precautions are taken to check for valve leakage.
A. Operability of Affected Penetration Assembly An operational check found that a leak exists at Valve 955 which, has been quantified to be 300 ml/min (Ref. 3.24). It has been
GL 96-06 Response
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R. E. ~ Ginna Page 18 ascertained that if heatup occurs in 10 seconds from ambient to 286 F, a relief rate of 215 ml/min will avoid thermal overpressurization in the penetration assembly. Since the measured value of 300 ml/min is greater than the minimum required, the line will not overpressurize (Ref. 3.25).
B. Action Planned For Permanent Protection Of Penetration Assembly A PCR is currently being prepared to install a relief valve on the Hot Leg Sample System. Installation is planned during the next refueling outage since the PCR will involve welding inside containment for supporting the relief valve.
2.3.2.3 Penetration 206a
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- Pressurizer Liquid Sample Review of PAID (Ref. 3.21) and Figure 6.2-53 of Reference 3.20 revealed that pressurizer liquid sample'can get trapped in the penetration assembly. Reference 3.20 showed that assuming Valve 953 does not leak, the high temperature fluid (higher than 286 F) will contract when cooled to ambient temperature and when it gets reheated to 286 F will not fill the line completely, thereby avoiding thermal overpressurization.
However, with a postulated small leak at Valve 953, the assembly'an subsequently get filled with fluid at ambient temperature which may overpressurize the assembly during design-basis accident conditions.
A. Operability of Affected Penetration Assembly An operational check found that a leak exists at Valve 953 which has been quantified to be 1500 ml/min (Ref. 3.24). It has been
4 a GL 96-06 Response R. E. Ginna Page 19 ascertained that if heatup occurs in 10 seconds from ambient to 286 F, a relief rate of 215 ml/min will avoid thermal overpressurization in the penetration assembly. Since the measured value of 1500 ml/min is greater than the minimum required, the line will not overpressurize (Ref. 3 25).
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B. Action Planned For Permanent Protection Of Penetration Assembly A PCR is being prepared to install a relief valve on the Pressurizer Liquid Sample System. Installation is planned during the next refueling outage since the PCR will involve welding inside containment for seismic support of the relief valve.
2.3.2.4
~ ~ ~ Penetration 207a - Pressurizer Steam Sample Review of PAID (Ref. 3.21) and Figure 6.2-55 of Reference 3.12 revealed that steam can get trapped between Valves 951 and 966A.
If Valve 951 does not leak, there is no concern because the assembly will not become water solid since the fluid trapped is steam.
However, if Valve 951 is leaking, the assembly can become full of ambient temperature liquid which may cause thermal overpressurization if the leak is small enough that it cannot relieve the pressure buildup.
A. Operability Of Affected Penetration Assembly An operational check found that indeed valve 951 does not leak (Ref. 3.24), thereby confirming the original finding that the line will.
not overpressurize.
1 B. Action Planned For Permanent Protection Of Penetration Assembly A PCR is being written to install a relief valve on the Pressurizer
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GL 96-06 Response R. E. Ginna Page 20 Steam Sample System. Installation is planned during the next refueling outage since the PCR will involve welding inside containment for seismic supports of the relief valve. In the interim, the valve will be monitored to ensure that it remains leakproof.
2.3.2.5 Penetration 307 - Fire Service Water Review of PAID (Ref. 3.22) and Figure 6.2-64 of Reference 3.12 revealed that cold water can get trapped in the fire service water system and jeopardize the penetration assembly during design-basis accident conditions due to thermal overpressurization. Reference 3.20 showed that both the assembly and the process line may potenially rupture due to high fluid pressure.
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A. Operability Of Affected Penetration Assembly
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To assure operability of the Penetration 307, a temporary precaution was implemented to either drain the water out of the assembly or open a vent valve inside containment (i.e., Valve 9255).
Either of these actions will not affect system performance when needed.
B. Action Planned For Permanent Protection Of Penetration Assembly A PCR is being prepared to install a relief.valve inside containment to relieve fluid pressure during design-basis accident conditions.
Installation can be done during Ginna normal operation mode.
2.3.2.6 Penetration 324 - Demineralized Water Review of PAID (Ref. 3.23) and Figure 6.2-73 of Ref. 3.12 showed that cold water may get trapped between Valves 8422 and CIV
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Page 21 8418. Design Analysis (Ref. 3.20) showed that the demineralized water line will rupture at a lower pressure than the penetration assembly thereby eliminating the concern of thermal overpressurization for the assembly.
A. Operability Of Affected Penetration Assembly The demineralized water pipe is made of seamless drawn copper, while the penetration assembly is carbon steel. As mentioned above, concern for thermal overpressurization of the penetration assembly is eliminated, assuring operability of the component.
B. Action Planned For Permanent Protection Of Penetration Assembly A PCR is being prepared to install a relief valve in the system to relieve potential thermal overpressurization in both demineralized water line and the penetration assembly. Installation can be done during Ginna normal operation mode.
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Page 22 2.3.2.7 Summary - Overpressurization Of Isolated Piping Review of PS.lD's of all safety related systems inside containment that are needed to mitigate design-basis accident conditions showed several ones that may be subject to thermal overpressurization.
Review included all containment penetrations and non-safety related systems that could interact with the safety related ones. These
-findings are su'mmarized below.
System Affected Corrective Action Status Temporary Permanent Containment Spray Filter Isolate & Install Relief Valve Dousing Line Vent Relief installed 1/3/97 I
Valve Penetration 1 21 a-Press. Valve 548 Install PCR is being Relief Tank Makeup Water Relief Relief prepared Valve Penetration 205-RCS Loop Valve 955 Install RV to be installed B Hot Leg Sample Leakage Relief next refueling Valve outage Penetration 206a-Press Valve 953 Install RV to be installed Liquid Sample Leakage Relief next refueling Valve outage Penetration 207a-Press Not required Install RV to be installed Relief next refueling Valve outage Penetration 307-Fire Drain or open Install RV to be installed Service Water vent Relief prior to outage Valve Penetration 324- Not required Install RV to be installed Demineralized Water Relief prior to outage Valve
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Page 23 3.0 List Of References 3 ~ 1 Generic Letter 96-06, "Assurance Of Equipment Operability And Containment Integrity During Design-Basis Accident Conditions", received by RGB.E on October 2, 1996.
3.2 EWR 5275, "Containment Fan Cooler Coil Replacement", 1991
- RGB.E Dwg. No. 21489-0657, "Containment Recirc. Fan Cooler "A" Enclosure-EWR 5275, Rev. 3, 1/21/94
- RGRE Dwg. No. 21489-0658,0659, 5. 0660, "Containment Recirc. Fan Cooler "B", "C", "D" Enclosure-EWR 5275, Rev. 1, 1/21/94 3.3 Dominion Engineering, Inc. (DEI) Design Report, DEI 356, Rev.
1, "EWR 5275 Containment Recirculating Fan Cooler Replacement".
3.4 RGBcE PAID Drawing 33013-1250, Shts 1',2 5. 3, "Station Service Cooling Water (Safety Related)", Rev. 14, 7/19/96.
3.5 GAI Drawings C-381-358 Shts. 9 (Rev.1), 13 (Rev.3), 18 (Rev.
1), 19 (Rev. 2), 20 (Rev.1), 21 (Rev. 2), 22 (Rev. 2), 23 (Rev.
1), 24 (Rev. 1), 25 (Rev. 1).
3.6 NUREG/CR 5220, "Diagnosis Of Condensation-Induced Water-hammer", Vol. 1 5. 2, dated October 1988.
3.7 EPRI Interim Report, "Containment Air Cooler Heat Transfer
'uring Loss Of Coolant Accident With Loss Of Offsite Power Conditions", Noy. 1996, Prepared By Numerical Applications, Inc.
3.8 EPRI Draft Report, "GOTHIC Analysis Of A Containment Fan Cooler Unit And Piping Under LOCA And Loss Of Offsite Power
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GL 96-06 Response R. E. Ginna Page 24 Conditions", May 1996, Prepared By Numerical Applications, Inc.
3.9 RGRE Design Analysis, DA-ME-96-073, "Containment Recirc Fan Cooler Water Hammer Evaluation", 8/14/96 3.10 RGSE Design Analysis, DA-ME-94-012, "Boiling In The SWS Discharge Piping Of The CRFC's", Rev. 0, January 22, 1994.
3 11 Sections 6.1, 6 2.2 And 9.4, R. E. Ginna Nuclear Power Plant
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UFSAR, Rev. 10 3.12 Ginna UFSAR, Sections 6.2.4, 6.6 Including Figures 6.2-13 to 6.2-78 Inclusive. Also Table 6.2-15.
3.13 RG5E PAID Drawing No. 33013-1261, "Containment Spray",
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Rev. 25.
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3.14 Ginna UFSAR, Figures 6.1-1 5 6.2-2 3.15 RGBcE Design Analysis, DA-ME-96-117, "Thermal Pressurization Of Containment Charcoal Filter Dousing Header", Dec. 20, 1 996.
3.16 RGSE LER 96-015, January 22, 1997.
3.17 RG&.E PCR 96-128 & Safety Evaluation, SEV 1088, 12/29/96.
3.18 RGRE Action Report 96-1250 3.19 RGSE PAID Drawing 33013-1258, Reactor Coolant Pressurizer", Rev. 17.
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~ Ginna Page 25 3.20 RG &E Design Analysis, DA-ME-96-119, "Thermal Pressurization Of Containment Penetrations 121a, 205, 206a, 207a, 307 & 324", 12/21/96.
3.21 RG&E P&ID Drawing 33013-1278, Sht. 1, Rev.13.
3.22 RG&E P&ID Drawing 33013-1991, Rev. 11.
3.23 RG&E P&ID Drawing 33013-1908, Sht. 3, Rev. 8.
3.24 Data From Gregg Joss (Performance Monitoring Section),
January 22, 1997.
3.25 NRC Meeting Minutes, "Meeting With NEI and Licensees to Discuss GL 96-06", November 22, 1996.
3 26 Marlo Coil Drawing No.
~ 17917C050, "Coil Assembly Details",
Rev. D, 2/18/93.
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Page 26 4.0 LIST OF ATTACHMENTS 4.1 Service Water Inlet And Outlet Piping For CRFC "A" (Schematic) 4.2 Service Water Inlet And Outlet Piping For CRFC "B" (Schematic) 4.3 Service Water Inlet And Outlet Piping For CRFC "C" (Schematic) 4.4 Service Water Inlet And Outlet Piping For CRFC "D" (Schematic) 4.5 Schematic Showing Service Water Inlet And Outlet Piping For A Typical Containment Recirculation Fan Unit.
4.6 GL 96-06 Evaluation Of Penetration Piping Systems For Thermally Induced Overpressurization.
RESPONSE TO GENERIC LETTER 96-06 ATTACHMENTS
ATTACHMENT 4.1 SERVICE WATER INLET AND OUTLET PIPING FOR CRFC "A" (SCHEMATIC) 18"5 & 14"5 SERVICE WATER DISCHARGE HEADER 14 0 SERVICE WATER HEADER EL 264'6" EL261'8" EL.262'3 EL262'9" CRFC "A" EL.256'3" EL.258'S'4 TO DISCHARGE P398 TUNNEL P319
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ATTACHMENT 4.2 SERVICE WATER
~ ~ ' INLET AND OUTLET PIPING FOR CRFC "8" (SCHEMATIC) 18 0 8c 144 SERVICE WATER
~ g DISCHARGE HEADER 14"0 SERVICE EL.263'6" WATER HEADER EL.263'3" EL.264'9" EL.262'3 3/8" EL.262'3" EL.264'6'RFC EL.262'9" "8"
EL258'3" TO EL.256'3" DISCHARGE P311 TUNNEL P316
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ATTACHMENT 4.3 SERVICE WATER
.INLET AND OUTLET 4 ~
PIPING FOR P
CRFC "C" (SCHEMATIC)
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18'8 dc 14"0 SERVICE WATER DISCHARGE HEADER 1 ~
14O SERVICE EL.
WATER HEADER 266'L- EL. 264'6" EL. 263'3" 267'3" 4" EL. 262'3" EL. 268'2 1 EL. 262'9" CRFC "C" TO EL. 258'3" EL. 258'S'315 DISCHARGE TUNNEL P328
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ATTACHMENT 4.4 SERVICE WATER INLET AND OUTLET PIPIN G FOR CRFC D (SCHEMATIC) 18 4 dc 14"O SERVICE WATER DISCHARGE HEADER 14"0 SERVICE EL. 265'6 264'6" EL.
WATER HEADER EL. 267'3 EL 262 5 3 4 EL 263 EL. 262'9 CRFC "D" 258'3" TO EL.
EL 256'3" DISCHARGE P323 TUNNEL P312
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AlTACHMENT 4.5 SCHEMATIC SHOWING SERVICE WATER INLET AND OUTLET PIPING FOR A 1YPICAL CONTAINMENT RECIRCULATION FAN UNIT