ML20199G719
| ML20199G719 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 01/28/1997 |
| From: | Graves J ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20199G490 | List: |
| References | |
| GGNS-97--R, GGNS-97-000-R00, GGNS-97-0002, GGNS-97-2, GL-96-06, GL-96-6, NUDOCS 9711250264 | |
| Download: ML20199G719 (53) | |
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Engineering Report NI GGNS.97 0002
,. Page lof_22
. .eision: o GRAND GULF NUCLEAR STATION ENGINEERING REPORT GGNS-97-0002, Rev. O FOR Generic Letter %06 Eva:uation of Drywell and Containment Penetrations l\ .
Prepared By: J. D. Graves . ~ t v- D te: ilte/9 7
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Reviewed By: /I/ (( .
- /44- Date: I/2 f[7 7 CGS / Reviewer Approved By:
' Date: I!28k7
/ /
ResponsQager 7-9711250264 971120 PDR .ADOCK 05000416 j P PDR.
- c. .n i
l Wag Rapn No: GGNS4976 l
Pap NJ1 i Revision: 9
, J Table of Contents l t
i Page l Description t
3
- Discussion !
Evaluation !
Drywell Cooling / Chilled Water System Evaluation 4 Penetration Overpressuration Evaluation 6 l 9
Refereos i Tables Table 1: Containmnet and Drywell Penetration Screening Results 11. l Table 2: Penetrations Susceptible to Overpressurization 21 i 22 !
Table 3: Penetrations Predicted Maximum Delta T Results 1
Attachments
-Attachment 1: Summary of 96 06 Penetration Structural Calculations Attachment 2: Safety Evaluation Applicability Review l t
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Engineering Report No: GGNS 9Mn02 Page lof 21 RevisionQ DISCUSSION:
On September 30,1996, the Nuclear Regulatory Commission issued Generic Letter 96-06 entitled Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions. The letter pertained to three items of concern that the NRC identified as generic implications from recent events at other nuclear plants.
The first item was identified in February 1996 by Pacific Gas and Electric (Diablo Canycn Unit I snd 2) to be a waterhammer event on the containment cooling water piping that supplies water to containment air coolers. PG&E determined that during a design basis LOCA with a concurrent LOOP, the cooling water pumps and the air cooler fans would lose power. Upon establishing ESF power, the air cooler fans would start, but full containment cooling water flow across the coils would not be established for approximately 30 seconds. During this interim, high temperature containment air would be forced over the coils and create a substantial steam volume.
When flow is re established, the pumped liquid may rapidly condense this steam volume and produce a waterhammer. The hydrodynamic loads introduced by such a waterhammer event could be substantial, challenging the integrity and function of the containment air coolers and associated component cooling water system.
The second item ofinterest involved the peutial for two-phase flow in safety-related piping and components. In an inspection of Connecticut Yankee Atomic Power Company (liaddam Neck-Nuclear Power Plant) in July 1996, the NRC identified an issue relative to two-phase flow in the station service water system supplying the containment coolers during design-basis accident conditions. In particular, the licensee's service water system pipe model did not consider two-phase flow through the cooling coils. This two phase mixture would result in a higher pressure
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drop in the service water piping and would affect the fluid flow and heat removal capabilities of
( the coolers during accident conditions. .
The final area of concern was identified by Dequesne Light Company (Beaver Valley Unit 1 & 2) in July 1996. During a surveillance test of a component cooling water inlet valve to the RIIR bcat exchanger on Unit 1, tlie motor operated butterfly valve located inside the containment would not open. They found tnat pressure in the piping section between this valve and a closed manual butterliv valve located outside the containment measured slightly higher than 'he system design pressure. The licensee concluded that pressure in the isolated section of piping increased when the trapped water was heated up by increased ambient temperatures.
The NRC has requested G01J to evaluate the two following questions pertaining to the three areas of concern noted above.
1.) Are containment air coolers cooling water systems susceptible to either waterharrmer or two-phase flow conditions duting postula'ed accident conditions?
d 4 Engineenng Report No: GGNS 97 n00J Page iof 22 Revision: o 2.) Are piping systems that penetrate the containment susceptible to thermal expansion of fluid so that over pressurization of piping could occur? The thermal expansion may be due to an accident, seasonal temperature variations, or normal temperature variations from start-up/ shutdown.
GGNS provided a 30 day response to the NRC on October 30,1996 to inc'icate (1) whether or not the requested actions will be completed, (2) whether or not the requested information will be submitted and (3) whether or not the requested information will be submitted within the requested time period. The NRC requested for all utilities to submit within 120 days a written summary report stating actions taken in response to the sequested actions noted above, conclusions that were reached sciative to susceptibility for waterhammer and two-phase flow in the containment air cooler cooling water system and overpressurization of piping that penetrates containment, the basis for continued operability of affected systems and cornponents as applicable, and corrective actions that were implemented or are planned to be implemented. This report documents engineering evaluations performed in response to the Generic Letter.
EVALUATION:
1.0 Drywell Cooling / Chilled Water System Evaluation The purpose of the Drywell Cooling System is to provide cooling air throughout the Drywell during the full range of normal operating conditions. The Drywell Cooling System consists of 6 cooler units located in the Drywell at elevations 161',147', and 93'. Each of the six coolers has two 100% capacity vane axial fans and two 100%
capacity cooling coils with only one of each normally in senice. The fans discharge
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into a common outlet plenum through separate discharge dampers. The dampers are
)' pneumatically operated and one closes upon loss of power / air while the other damper opens. Each fan and its asso::iated discharge damper solenoids in.one train of the Drywell' cooling system are powered from an ESF bus to provide for operation during a LOOP.
The Drywell Chilled Water System is a closed loop system consisting of four 50%
capacity chillers and two 100% capacity centrifugal pumps located in the Auxiliary Building at el:,ation 119', expansion tank located in the Auxiliary Building at elevation 185', and other associated equipment necessary to support system operation.
The purpose of the system is to remove the heat absorbed by the Drywd Cooling System and transfer it to the Plant Senice Water System. One of the chilled water pumps and two of the chiller units are supplied from an ESF bus to provide for operation during a LOOP. Additionally, chiller cooling water can be supplied manually from Standby Senice Water in support of LOOP operations. The system is maintained full by use of an expansion tank with level automatically maintained with water from the Makeup Water Treatment System. The Drywell Chilled Water System i
- Enginecting Repn Nr GGNS 97 0002 Page 1 of 22 Revision: 0 is equipped with two sets ofisolations, one set for supply water (IP72F124/147) and one set for return water, valves (IP72F125/126) that provide for system isolation in the event of a LOCA.
As stated in the UFSAR Section 9.2.11.3, the Drywell Chilled Water System is non-safety related except for its isolation valves. Failure of the system will not compromise any safety related system or component and will not prevent safe reactor shutdown.
The system is provided with features to allow restoration and use following a LOOP and GGNS Emergency Precedures direct operations to return the system to operation any time the Drywell temperature exceeds 135T Generic Letter Question 1:
Are containment air cooler cooling water systems susceptible to either waterhammer or two-phase flow conditions during postulated accident conditions?
This question is not applicable to GGNS since the Drywell Coolers / Chilled Water System are non safety related except for its isolation valves. Although the susceptibility of this system to the prer,cribed water hammer has not been explicitly evaluated, failure of the system due to a water hammer event will not compromise any safety-related system or component and will not prevent safe reactor shutdown.
Should the system be retumed to operation following an isolation after a LOCA and a water hammer occur, the loss of chilled water inventory would be annunciated on a control room panel (H13P870-4A) when FAL L600A/B detects a low discharge pressure and the chilled water pumps trip. Existing procedures are in place to recognize the event and close isolation valves IP72F124/147 and IP72F125/126 to establish containment.
Similarly, should the system be retumed to service following a LOCA and two phase flow through the cooler occur, it is not applicable at GGNS since loss or failure of the system or its ability to r move heat from the Drywell will not compromite any safety-related system or component and will not prevent safe reactor shutdown. The loss of heat transfer from the cooling coil due to 2 phase flow would be insignificant since no creilit is taken for coolen, during accident mitigation. Similarly, should the 2 phase flow initiate a water hammer in the cooling coil, its breach would be handled as described in the preceding paragraph.
Containment coolers were not evaluateo :ince they are isolated on a LOCA signal and are not retumed to service. Additionally, the Containment em>ironment during ..
LOCA reaches a maximum of 181T, which is below the anticipated saturation temperature of the water in the coils. Consequently, a evaluatic,n of a water hammer or two phase flow event on Containment Cooler was not required. Reduction of the saturation temperature of the water in the coils doe to a particle vacuum caused from
Engineering Report No: GGNS 97.n002
.- Page6of11 Revision:,Q dreie W ek of the chilled wat is not anticipated since the isolation valves are assumed to beak tight for overpressurintion evaluation.
2.0 Penetration Overpressuration Evaluation NPE performed the following setions to analyze the impact of overpressuration of penetrations due is thermal expansion.
1.) NPE identified the Drywell and contairunent penetrations that would be considered for evaluation, see Table 1. These pc'tetrations were subjected to the following screening criteria to determine if any ferther analysis was required. It should be noted that all valving is considered to be in its operation position ar stated I. .he FSAR Table 6.2-44, Containment Isolation Valve Information. Off normal valve positioning due to maintenance or testing is not considered to be credible. A penetration would be considered to be affected by overpressuration ifit meets gjl of the following criteria:
A.)The penetration must be full of liquid at the time of the accident. Pipes containing air, gas or steam will be excluded B.) The liquid contained in the penetration piping must be at a lower temperature than the surrounding erwironment during operational or accident situations.
Piping that contains water at or near RPV or RWCU temperatures would actually have initial fluid temperatures higher than those expected during an accident.
C.) The penetration must be isolated during an event, i.e. plant heatup or accident, l
that could cause a significant heat transfer to the fluid between the isolation j'
valves. The valve arrangement used for penetration isolation must restrict flow out both directions. If the inboard isolation vawe is a cLk valve or certain type and orientation of solenoid valve, (with a mechanism of pressure relief on the connecthg piping) the penetration may possibly be excluded. This exclusion would also include piping open to the suppressior, pool, RPV, or containment air space.
In order to be excluded, the extended piping system available for fluid expansion inside containment must not constitute a closed system, so that the fluid volume can expand and prevent damage to the containment isolation portion of the piping penetration.
Additionally, another closed valve further down he line inside containment mt'st not prevent expansion of the fluid volume in the penetration, thereby isolating a penetration with an expected available leak path i.e. check valve.
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I Engineering Repon No: GGNS-974002 )
Page 2 of 22 d
Revis;on:,0 D.)The potentially susceptible penetration will not have any pressure relief valves (with sufficient capacity and setpoint) or other method of overpresmre protection (such as a check valve in parallel with the main inboard valve) between the isolation valves.
A penetration will additionally be considered potentially susceptiMc ifit meets any one of the following two criteria:
The penetratic>n will be considered potentially susceptible if a loss of power event coupled with an accident would cause isolation, heatup and overpressurization, outside design requiremerts, of a normally open, low temperature, Duid fdled penetration.
A penetration will be considered potentially susceptible if trapped pressure can prevent safety related isolation valves from opening when required to mitigate an accident, i.e. pressure locking. Ref. Generic Letter 95-07.
A potentially susceptible penetration may be eliminated from concern if qualified calculations or analysis demonstrate that the penetration piping system, which includes the valves, remains within its 1)esign requirements, i.e.
ASME Code allowable stresses for faulted conditions.
Table I contains the results of the screening and an explanation of the results.
Eighteen penetrations require further evaluation.
2.) For penetrations that are affected by overpressurization, NPE will determine the stress the pipe is subjected to due to expansion of the trapped water if the stiess
( is less than ASME Code, the. penetration may be excluded pending further evaluation. ~.
NPE completed Calculation MC-Qllll-97001, Rev. O to determine the approximate maximum temperature and pressure of the water between containment isolation valves for specific penetrations. The penetrations susceptible to overpressuration are listed in Table 2 and are subject tc, the following general assumptions.
A.) Nominal pipe dimensions are used throughout the calculations.
B.) The maximum temperature is based on the physical layout of the piping system as shown in Engineering Calculation MC-Qllll-97001, Rev. 0. The piping was modeled as a horizontal pipe in a convective emironment oflaminar air DoW.
C.)The atmosphere in the Drywell for the first 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of the accident will be 330
- F, superheated steam. This temperature occurs for a small line break event,
Engmeering Report No: GGNS 97 6002 Page1oQ2 Raisba:.Q see FSAR Table 3.113, Rev.10, Accident Environment Inside Drywell, Containment, and Auxiliary Building, and GGNS E-100.0, Rev. 5, Curve Set 1.
D.)The Containment temperature profile for a small break LOCA can be found in FSAR Figure 6.2 97, Rev. O. While this is actually a temperature profile for the suppression pool, it will approximate the Containment temperature.
E.) The temperature profile for the Auxiliary Building room where the penetrations are located will not be affected by a small break LOCA in the Drywell since little operator action is assumed to occur during the first six hours of the accident, FSAR 6.2.1.1.3.3.5. The only reject of Containment heat that could change the Auxiliary Building temperature would be operation of I train of RIIR in suppression pool cooling mode. This would provide some localized heatup in the RHR room for the train in operation, but would not P.frect other rooms. If the train in operation should be train 'A', penettstion 47 couM be expected to experience a teniperature of 124 *F in room 1 A203. Otherwise, the Auxiliary Building room teenperatures will be assumed to be the normal temperature profiles listed in GGNS E-100.0, Rev. 5 for the room where the penetration enters Con'ainment.
The results of the calculations indicate that the 18 penetrations identified in Table 2 and Attachment I have the potential to exceed the ASME code allowable stresses bases on the maximum expected final temperatures and pressures.
3.) Should a penetration affected by pressure locking be subjected to a stresses that have the potential to exceed the ASME code allowable limits, NPE will have Component Engineering perform an analysis in accordance with ,uirements in ASME, Section NA, Appendix F. and the criteria contained in A ament A.
h Additional evaluations of these penetrations have been performed based on heat transfer (Calculation MC-Q1111_97001, Rev. 0) and Finite Element Analysis methodology (Calculation QR-079-07, Rev. 0), in an effort to satisfy the operability requirements of GL 91-18. The purpose of the finite element analysis (FEA) was to estimate conservative temperature increases which would be required to cause the failure of various size penetrations. The heat transfer calculation was used to develop the approximate temperature increase that would be expected for the penetrations. In general, the results of the heat transfer calculation showed that the anticipated temperature increases were small in relation to the temperature increases expected for penetration failures. Based on these calculations and engineering judgment, none of the penettstions are expected to experience temperatures which would lead to a penetration failure. The penetrations are not expected to fait during the bounding accident thus no challenge to contairunent integrityis expected.
Engineering Repon No: QgNS.97 0002 Page 2 of n Revision:,Q 4.) If u penetration effected by thermal expansion is not excluded by items 1 through 3, NPF will evaluate its failure on the continued operation of the reactor.
In order to provide additional conservatism, NPE has evaluated the effects of a postulated failure of penetration 331, which exhibited the worst case heatup. In addition to the penetration failure, an additional single active failure of one isolation valve (i.e. Opens) would occur to result in a leakage path from the Drywell to Containment. Penetration 331 is part of the Drywell boundary which would limit the amount of Drywellleakage bypassing the suppression. Procedural limits imposed on Surveillance testing for Drywell bypass leakage ensures that the actual value ofleakage is a small fraction of the design limit. The actual leakage based on this margin between the Surveillance limit and the design limit bounds any single opening in the Drywell of 6 inches or less. Continued ability of the Drywell to perform its safety related function is assured since Penetration #331 has a nominal inner diameter of approximately 4 inches and its leakage is bounded by this margin.
REFERENCE:
1.) SD h151/P72, Drywell Cooling Sy: tem, System Description, Rev. 0 2.) FSAR Section 9.2.11, Drywell Chilled Water System, Rev. 0 3.) Engineering Report GGNS-94-0052, Evaluation of Containment Leak Paths 4.) Technical Requirements hianual Table TR3.6.1.3-1, Primary Containment Valves, R;:v.13
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I' ?cchnical Requirements hianual Table TR3.6.4.2-1, Secondary Containment Automatic 5.)
Isolation Valve's, Rev.13
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6.) Technical Reg: irements hianual Table TR3.6.5.3 1, Dryweil isolation Valves, Rev. 6 7.) FSAR Table 6.2.44, Containment Isolation Valve Information, Rev. 8 8.) Drawing 9645-C-1063, Unit I Containment Drywell Wall Penetrations-Schedule and Shielding Details, Rev.17 9.) Drawing 9645-C-1003, Unit 1 Containment Cylinder Wall Penet. Schedule & Radiation Shielding Detalla, Rev. I1 10.) F3 AR Table 6.2-44, Containment Irolation Valve Information, Rev. 9
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. . . 5 Empasering Raport No: GGNS997 0001
. i Pass 19ofn- j Revis6on:g .
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11.) FSAR Table 3.11-3, Accident Environmeat Inside Drywell, Containment, and Auxiliny i
Building, , Rev.10 r i
12.) GGNS E-100.0, Curve Set 1, Rev. 5 t 13.)' FSAR Figure 6.2-97, Small Break Accident (1 RHR) Event 3a, Rev. 0 14.)' FSAR Section 6.2.1.1.3.3.5, Small Size Breaks, Rev. 0 - ;
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e Engineering Repon N): GGNS 97 0002 Page H of H Revision @
Table 1 <
Containment and Drywell Penetrations Screening Results Penetration # Concern Explanation 4 No Tube is not subject to heating since it is below the level of the spent fuel / upper containment pool during LOCA.
5 No Working Fluid is steam 6 No Working Fluid is steam 7 No Working Fluid is steam 8 No Working Fluid is steam 9 No The Feedwater system would be operating at temperatures in excess of Drywell accident temperatures. Since the Drywell accident temperature is less than this temperature, the pipes would actually cool following an isolation.
10 No The Feedwater system would be operating at temperatures in excess of Drywell accident temperatures. Since the Drywell accident temperature is less than this temperature, the pipes would actually cool following an isolation.
I1 No Only one isolation valve. The other end of pipe is below the level of Suppression Pool No Only one isolation valve. The other end o r pipe is below the level of 12 Suppression Pool 13 No Only one isolation valve. The other end of pipe is below the level of
!, Suppression Pool 14 No Check valve IE12F308 relieves pressure between IE12F009B and h IE12F008A to the Drywell side of the inboard isolation valve. Piping downstream of F009B is open to the RPV.
15 No spare 16 No spare 17 No Working Fluid is steam 18 No The penetration is contained within a insulated guard pipe and it not subject to the Drywell/CTMT environment 19 No The penetration is contained within a insulated guard pipe and it not subject to the Drywell/CTMT erwironment
Engineering Report No: GGNS.97 0002 Pmudu Revision: 0,,
Tabie 1 Containment and Drywell Penetrations Screening Results Penetration # Concern Explanation 20 No The branching line between E12F027A (outside containment) and E12F02bA(Spray supply) and E12F042A(LPCI supply) is protected by a PV E12F025A, which is .et @ $00 psi.
21 No The branching line between E12F027B (outside containment) and E12F028B(Spray supply) and E12F042B(LPCI supply) is protected by a PV E12F025B, which is set @ 500 psi.
22 ho Drywell isolation valv1 E12F041C (Drywell) is a testable check valve and pressure would not be able to build between it and IE12F042C(Auxillary Bld.). There is no isolation valve in Containment valve.
23 No IE12F064A is the only isolation valve. The other end of pipe 4"-GBB.
37/4"-GBB-10 connects to 18"-HBB-32/12" HBB 76 terminates below the level of Suppression Pool 1 24 No IE12F064C is the only isolation valve. The other end of pipe 4"-GBB-72 connects to 18"-ISB-75 terminates below the level of Suppression Pool 25 No IE22F015C is the only isolction valve. The other end of pipe 20"-HBB-21 terminates below the level of Suppression Pool.
26 No Drywell isolation valve E22F005(Drywell) is a testable check valve and j pressure would not be able to build between i: and IE22F004(Auxillary Bld.). There is no isolation valve in Containment valve.
27 No IE22F012 is the.onlylsolation valve. The other ead of pipe 14"-HBB-32 terminates below the level of Suppression Pool.
28 No IE51F031 is the bnly isolation valve. The other end of pipe 6"-HBB-49 terminates below the level of Suppression Pool.
29 No IE51F063 is the only isolation valve. The other end of pipe 18"-IIBB 75 terminates below the level of Suppression Pool.
30 No IE21F001 is the only isolation valve. The other end of pipe 20"-HBB 8 tenninates below the level of Suppression Pool.
31 No Drywell isolation valve E21F006(Drywell) is a testable check valve and pressure would not be able to build between it and IE21F005(Auxillary Bld.). There is no isolation valve in Containment valve..
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Engin:4 ring Report No: GGNSo97 0002 Pass u oty Revision:Q l Table 1 Containment and Drywell Penetrations Screening Results ,
Penetration # Concern Explanation 32 No IE21F011 and lE21F012 are the only isolation valve. The other end of pipe 4"-GBB ll/14"-GBB 14 connects to 14"-HBB-9 tenninates below the level of Suppression Pool 33 No Containment isolation valve ICllF122 is a testable check valve and pressure would not be able to build between it and ICIIF083A. Relief valve 1ClIF025A would relieve pressure inside CTMT.
34 No Working Fluid in Air 35 No Working Fluid is Air 36 Yes Valves IP72F122 and F123 Close on LOCA signal and trap water between valves 37 No Containment isolation valve IP72F165 is a testable check valve and pressure would not be able to build between it and IP72F121A.
IP72F106 is first active RV tlist would relieve pressure downstream of F165. ~
38 No Containment isolation valve IP71F151 is less that 6" from the Containment Wall. There will be no area open for heat transfer in Containment. Therefore pipe will not heat up due to LOCA 39 Yes Valves IP71F148 and 149 close on a LOCA signal ano trap water between them.
40 No Working fluid is air l
41 No Working fluid is air
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42 No Working fluid is air 43 Yes Vr.lves 1G33F034 and F028 close on a LOCA signal and trap water between them. Since the valves are normally closed during operations, t
they would not be at a temperature comparable with RPV co: 0 temperature prior to isolation 44 No Containment isolation valve IP42F035 is a testable check valve and pressure would not be able to build between it and IP42F066A. Piping downstream ofF035 relieves by F224.
45 No Valves don't close on a LOCA 46 No: IE51F019 is the only isolation valve. The other end of pipe 4"-GB5-l 11/2" HBB-49 terminates below the level of Suppression Pool 47 Yes Valves IB33F127 and F128 are normally closed with a volume of water trapped between them i ,. .; _
Engineering Report No: GGNS,97 o002 Pagej{of22 Revision:,Q Table 1 Containment and Drywell Pesetrations Screening Results Penetration # Concern Explanation 48 No IE12F073B is the only isolation valve. The other end of pipe 10".HBB.
64/10".11BB.79 termint.tes below the level of Suppression Pool
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49 Yes Valves 1G36F106 and F101 isolate on a LOCA signal with a volume of water trapped between them 50 Yes Vaives IP45F068 and F067 isolate on a LOCA signal with a volume of '
water trapped between them 51 Yes Valves IP45F061 and F062 isolate on a LOCA signal with a volume of water trapped between them 52 spare 53 spare 54 Yes Valves G41F053 and G41F201 are locked closed gate valves with a volum of water trapped between the.a 55 spare 56 No Containment isolation valve IPilF004 is a testable check valve and pressure would not be abt to build between it and IP11F075. Piping dowrstream of F075 relieves to upper CTMT pool.
57 No Containment isolation valve IG41F040 is a testable check valve and pressure would not be able to build between it and IG41F028 A. Piping downstream of F075 relieves to upper CTMT pool.
58 Yes Valves G41F029 and F044 isolate on a LOCA signal with fluid trapped between them.
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59 spare 60 No IP45F278 is the only isolation valve. The discharge pipe, 4".HBB.247, terminates inside containment.
61 No Containment isolation valve IC41F150 is locked closed. The only operaticn of the valve appears to be during LLRT. During this test, the volume between IC41F150 and F151 is drained and the LLRT is done with air. Since no other operation of the valve can be found, the space between the valves is assumed to be occupied by air.
62 No Not used 63 No Not used 64 spare 65 Working fluid is ir 66 Working fluid is air 1 ,
l Engineering Report No: GGNS 97-0002 Pmudu Revision:,Q Table 1 i
Containment and Drywell Penetrations Screening Results Penetration # Concern Explanation 67 No IE12F064B is the only isolation valve. The other end of pipe 4"-GBB.
37/4"-GBB-10 connects to 18" HBB-32/12" HBB-76 terminates below '
the level of Suppression Pool 68 spare 69 No IPlIF130 is the only isolation valve. The suction pipe terminates below the level of Suppression Pool.
70 No Working fluid is air 71A No IE21F018 is the only isolation valve. The discharge pipe terminates below the level of Suppression Pool 71B No IE12F025C is the only isolation valve. The discharge pipe,1" HBB 77 terminates below the level of Suppression Pool 72 spare 73 No IE12F036 is the only isolation valve. The discharge pipe, 6" HBB 78 terminates below the level of Suppression Pool 74 .- - . . spare -- _ .. - -- . - - _ . .
75 No Working fluid is air. Pipe terminates in air space inside CTMT 76 No IE12F005 is the only isolation valve. The discharge pij e,1" HBB 83 terminates below the level of Suppression Pool IE12F055A is the only isolation valve. The other end of pipe 10" HBB-77 No 84/10" HBB 79 terminttes below the level of Suppression Pool 78 spare h
79 spare ,
80 spare 81 Yes Valves IB33F125 and F126 are normally closed with a volume of water trapped between them 82 No Working fluid is air 83 No The RWCU system would be operating at temperature comparable with RPV core temperature prior to isolation. Since the Drywell accident temperature is less than this temperature, the pipes would actually cool following an isolation.
84 Yes Valves IP45F098 and F099 isolate on a LOCA signal with a volume of water trapped between them 85 No IP60F110B is the only isolation valve. The suction pipe terminates below the level of Suppression Pool.
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Engineering Repon No: @NS-910002 Page16 of 21 Revision:.0 l
Table 1 i Containment and Drywell Penetrations Screening Results l 1
Penetration # Concern Explanation 86 Yes Valves IP21F018 and F017 isolate on a T OCA signal with a volume of water trapped between them ,
87 No The RWCU system would be operating at temperature compuable with '
- RPV core temperature prior to isolation. Since the Drywell accident temperature is less than this temperature, the pipes would actually coel following an isolation.
38 No The RWCU system would be operating at temperature comparable with RPV core temperature prior to isolation. Since the Drywell accident temperature is less than this temperature, the pipes would actually cool following an isolation.
89 No Containment isolation valve IP41F169A is a testable check valve and pressure w s... mt be able to build between it and IP41F159A. Piping downstream of F0169A and 168A relieves to valve F041.
90 No Valves P41F168A/160A open on Rx Low Level or High Drywell
- -Pressure. Piping-downstream-of-F0169A and 168A relieves to valve F041. . . _ . - -
91 No Valves P41F168A/160A open on Rx Low Level or High Drywell Pressure. Piping downstream of F0169B and 168B relieves to valve F041.
92 No Containment isolation valve IP41F169B is a testable check valve and
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j pressure would not be able to build between it and IP41F.59B. Piping downstream of F0169B 2nd 1688 relieves to valve F041.
10lx No impulse line, Working fluid is air 102x No Impulse line, Woiking fluid is air 103x No Impulse line, Working fluid is air 104x No Impulse line, Working fluid is air 105x No Working fluid is air 106x No Working fluid is air 107x No Working fluid is air ,
108x No Working fluid is air 109x No Working fluid is air 110x No Impulse line, Working fluid is air 111 No Working fluid is air 113 No Impulse line, Working fluid is air 114 No Impulse line, Working fluid is air 115 No Impulse line, Working fluid is air
Engineering Repon No: GGNS 97 0002 PmDdu hwon: 0 .
Table 1 Containment and Drywell Penetrations Screening Results Penetration # Concern Explanation 116 No Impulse line, Working fluid is air 117 No Impulse line, Working fluid is air ,
11S No Impulse line, Working fluid is air 119 No Impulse line, Working fluid is air 120 No Impulse line, Working fluid is air 200xx No Electrical penetrations 301 No Working Fluid is steam 302 No Working Fluid is steam 303 No Working Fluid is steam 304 No Working Fluid is steam 305 No The Feedwater system would be operating at temperature comparable with RPV core temperature prior to isolation. Since the Drywell -
accident temperature is less than this temperature, the pipes would actually cool following an isolation.
306 No The Feedwater system would be operating at temperature comparable with RPV core temperature prior to isolation, Since the Drywell accident temperature is less than this temperature, the pipes would actually cool following an isolation.
307 No Valve E12F308 relieves around inboard isolation valve E12F009 308 spare 309 spare h
310 No Workink Fluid is steam 311 No The penetration is contained within a insulated guard pipe and it not subject to the Drywell/CTMT environment 312 No The penetration is contained within a insulated guard pipe and it not subject to the Dmvell/CTMT environment 313 No Drywellisolation valve E12F041 A is a testable check valve and pressure would not be able to build between it and IE12F042A. Piping downstream of the testable check col ; cts to the RPV.
314 No Drywell isolation valve E12F041B is a testable check valve and pressure would not be able to build between it and IE12F042B. Piping downstream of the testable check connects to the RPV.
_ ______._ _._ _ _ _.._ _._._ _ _ _ _ ~ . _
t Engineering Report No: GGNS 97 0002 _
Pass.ltof22 Revisiend! :
-Table 1 ;
Contaientent and Drywell Penetrations Screening Results Penetration # Concern Explanation-315 No Drywell isolation valve E12F041C (Drywell) is a testable check valve i and pressure would not be able to build between it and IE12F042C(Auxillary Bid.). There is no isolation valve in Containment valve. . Piping downstream of the testable check connects to the RPV. .
316 No Drywell isolation valve E22F005(Drywell)_ is a testable check valve and pressure v ould not be able to build between it and IE22F004(Auxillary !
Bid.). There is no isolation valve in Containment valve.
317 No Drywellisolation valve E21F006(Drywell) is a testable check valve and pressure would not be able to build between it and IE21F005(Auxillary ,
Bid.). There is no isolation valve in Containment valve. r 318 No CRD withdraw & insert lines. No isolation valves ;
319 No CRD withdraw & insert lines. No isolation valves No CRD withdraw & insert lines. No isolation valves :
320 321 No CRD withdraw & insert lines. No isolation valves 322 No spare ;
323 No spare 324 No- spare The RWCU system would be operating at temperature comparable with 325 No '
RPV core temperature prior to isolation. Since the Drywell accident
! temperature is less than this temperature, the pipes would actually cool following an isolation. .
],
- 326- ..
IB33F013A is a. testable check valve. It will relieve pressure toward the Drywell.
327 spare ;
328 IC41F007 is a testable check valve. It will relieve pressure toward the ;
Drvwell.
i 329 No IP42F115 is a testable check vstve. It will relieve pressure toward the Drywell. IP12F225 is first RV.
330 Yes IP42Fil6 and Fil7 will close and trap water between valves 331 Yes IP72F125 and F126 Close on LOCA signal _ and trap water between i
valves 332- No IP72F075 is a testable check valve, It will relieve pressure toward the :
Drywell. - IP72F045 is first RV. >
333 Yes Water is trapped between IB33F205 & F206 ,
334 No Equipment Hatch 335' No- Working fluid is air p
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Engineering Rep-rt No: GGNS 97 00n2
,. Page 12 of 22 Raision:,0 Table 1 Containment and Drywell Penetrations Screening Results Penetr4 tion # Concern Explanation 336 spare 337 No The RWCU system would be operating at temperature comparable with RPV core temperature prior to isolation. Since the Diywell accident temperature is less than this temperature, the pipes wotdd actually cool ,
following an isolation.
338 No Working fluid is air 339 No Working fluid is air 340 No Working fluid is air 341 No Working fluid is air 342 No Empty 343 No Empty
,344 spare 345 No Working fluid is air 346 No IB33F013B is a testable check valve. It will relieve pressure toward the Drywell.
347 No Working fluid is air 348 Yes Valves P45F009/F010 isolate on a LOCA signal with fluid is trappd between them.
349 Yes Valves P45F003/F004 isolate on a LOCA signal with fluid is trapped I between them.
350 Personnel Airlock h
351 No Working fluid is steam 352 . spare 353 spare 354 spare 355 spare 356 spare 357 No Working fluid is steam 358 spare 359 spare 360 spare 361 spare 362 No Working fluid is air 363 No Working fluid is air .
i I
Engineering Report No: GGNS-97-0002 4
- Page 29 ofII l RMaion:.0, j Table 1 Containment and Drywell Penetrations Screening Results Penetration # Concern Explanation 364 Yes Valves P45F096/F097 isolate on a LOCA signal with fluid is trapped between them.
spare l 365 360 No. The RWCU system would be operating at temperature comparable with RPV core temperature prior to isolation.' Since the Drywell accident temperature is less than this temperature, the pipes would actually cool ,
following an isolation.
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Engineering Repon No: GGNS.9%')00)2 Page 21of 31 I
Revism:,2 Table 2 Penetrations Susceptible to Overpressurisation Pene # Description Pipe Class Pressure Class 330 Comp- C#a= Water Retum 8-HBB 37 150#@500F '
331 Drywell Chilled Water Returv 4-HBB 42 150#@500F 333 CRWST to RPV 4 HBB 1!! 1504@500F 348 Drywell Equipment Drain 4 HBB 95 1504@$00F 349 Drywell Floor Drain 4 HBB 96 150#@500F 364 ChemicalWaste Sump Pump Dischaue 1.5 HCB 20 150#@500F 36 Drywell ChilluiWater Retum 4-HBB-40 150#@$00F 39 Plant Chilled Water Return 4-HBB-43 150#@500F 43 RWCU to Main Condenser 6-EBB 1 600#@850F 47 Post Accident Sample Line 3/4-DCB-50 900#@l000F 49 RWCU hWash Transfer Pump 4-HBB 152 150#@500F
$0 CTMT Equipment Drain 6-HBB 102 150#@500F
$1 CTMT Floor Drain 6-HBB 101 150#@500F-54 RWST to Upper CTMT Pool 12-HBB-4 150#@500f-58 Upper Pool to Fuel Pool Drain Tank 8 HBB-6 150#@$00F 81 Post AceWat Sampic Line 3/4 DCB 51 900#@l000F B4 ChemicalWas'.e 5 PunapIfW as e 3 HCB 19 150#@$00F 86 Dominer=1M Water Supply' ~~' 2-HBB-155 '" 150#@$00F f
1
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Engineering Report No: GGNS 97 6002
- Page H of H Revision:.9 Table 3 Penetrations Maximum Delta T Results Pene Predicted Volume Normal Predicted Delta-T
- Press (psi) ,
Int: ease % Temp (7) Max. Temp CT) (T) 330 5600 0.127254 105 167 ,
62 331 9950 0.234483 75 193 118 333 7740 0.146471 110 187 77 348 4240 0.085524 120 163 43 349 3400 0.066525 135 170 35 364 9180 0 253414 135 223 88 1670 0.057521- 50 94 44 36 39 3470 0.105037 65 132 67 1260 0.000030 120 131 11 43 47 5540 0.168394 90 161 71 49 2500 0.075919 100 143 43 50 1250 0.027002 120 138 18 51 2090 0.065114 105 133 28 54 70 0.000000 125 124 1 58 360 0.018001 90 99 9 81 2470 0.115544 90 140 50 84 1190 0.042006 135 153 18 4980 0.152477 65 139 74 86 ,
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Anachment 1 ' (
~
- Tc: EmpneenasReport No: GGNS 97 4002. Rev 0 ;
- Page l ef 3
. [
Summary of %06 Psmiion Structural Calculations Generic Letter 406 requires the evaluation of all drywell and containment penetrations for the effects of ;
t
- thermal volumetric exparsion of the water between the isolation valves.
i Eighteen of Grand Gulf's penetrations were identified as being susceptible to this scenano.
t I
- All eighteen of the susceptible penetrations are for non-safety related systems.-
in order to meet the Code allowables for this event (Design B: sis, as defined in MS-44), both of the a 4
i following requirements must be met: i The maximum pressure must be less than- twice the Design Pressure from MS-02, and l I
The maximum primary stress (from Paragraph NC-3652.2, ASME Section III,1974 Edition) during this event n.ast be leu than the Faulted stress allowable of 2.4 Sn. (Including pressure,
' deadweight, seismic, and hydrodynamic loadings) I
^
In order tojustify continued operation of the plant, Generic Letter 91-18 allows piping to exceed Coce allowables until the next scheduled outage, at which time the piping must be shown to be within Code ,
allowable or be brought within Code allowables by physical modification.. However, during the interim, the piping must meet operability allowabics, which are detailed in Appendix F of the ASME Code.
In order to show that the piping is operable, the following requirements must be met (Ref. Appendix F):
The maximum pressure :nust be less than twice the value obtained from Equation NB-3641.l(2) of the ASME Code,Section III,1974 Edition, and T _ i The maximum primary stress in the piping (as calculated in Paragraph NB-3652, ASME Section III,1974 Edition)must be less than 3.0S.._(Obtained using standard analytical. techniques)
OR:
A finite element analysis I can be performed, using an elastic-plastic methodology, and the ;
i maximum primary stress obtained by this method must be less than the greater of 3.0 S. or 0.7S..
Approximately half of the Grand Gulf penetrations meet the f.rst requirement,- but the other half can be
. qualified for operability only by resorting to the second method.-
~
The finite element analysis which has been performed is a simolified model which employs a circular cross section only. Two models were run , a ten inch, sch. 40 pipe and a 3/4 inch, sch 80 pipe, which is felt to bound the wide range of actual piping sizes. No modeling was done for some of the discontinuities which
- exist, such as reducers, elbows, branch connections, tapered transition joints, socket welds or valves.
Only pressure loading was considered.
4 i
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- Annachment 1 '
- To: s'agineering Repon No: GGNS-97-0002. Rev.0
' Page 2 of 3 It is assumed that no seismic event will occur during this scenario (based on a probability of 2 X E-9 that an earthquake will occur during this event)and that the only loads on the piping are deadweight and .
pressure.
Analytical results from the finite element equation can be obtained from Calculation QR-079 07, Rev. O.
The results are shown below:
For 10", sch 40 pipe, the maximum Delta T is 1147, max. stress = 52,300 psi; 3S,.= 60,000 psi.
For 3/4", sch 80 pipe, the maximum Delta T is 200T, max. stress = 46,400 psi; 35.= 60,000 psi.
The highest Delta T calculated for Grand Gulfis for a 4" penetration (#331) with a Delta T of 118 T.
Calculation QR-079-07 results predict that for smaller D./t ratios, the Delta T for failure, increases.
Therefore, since this penetration (#331) exceedance is only 4'F above the 114 'F for the 10 inch pipe, it is expected to maintain pressure integrity. Even if this penetration fails, however, the Design Basis leakage rate from drywell to containment is not exceeded. The next highest Delta T is about 88 'F for a small bore pipe, with an allowable of close to 200F and is not conaldered to fail. Also, enough margin exists to accommodate deadweight. The remainder of the penetrations have Delta T's much lower than the allowable Delta T's and, ti n fore, are considered to maintain pressure integrity and can be considered to be operable. Valves were not stress analyzed because typically they are considered stronger than pipe.
The results of the simplified TEA analysis are that all of the penetrations for Grand Gulf are operable and that the required safety functions are met.
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7t% Engineering Report No: OGNS 97 0002.Rev 0 Pase 3 of 3 l Twice Oper. Oper. Calculated Delta T Calculated :
Pen. Pipe i MS 02 Allow. Allow. Pressure Allowable Delta T No.
during (from l Design Pressure NB3641 NB3652 event FEA)
(Code Compl.)
250 - 22% 4106 5600 114 62 330 8 HBB 37 i 250 2871 5793 9950 114 118 331 4-HBB-42 5793 7740 114 77 333 4 HBB-lli 300 2871 2871 5793 4240 114 43 l 348 4-HBB-95 300 300 2871 5793 3400 114 35 349 4 HBB 96 300 5310 8719 9180 114 88 364 1.5 HCB-20 250 2871 5793 1670 114 44
-36 4 HBB-40 300 2871 5793 3470 114 67 39 4 HBB-43 6-EBB 1 2840 4439 7173 1260 114 11 43 2500 10855 13229 5540 200 71 47 3/4-DCB-50 300 2871 5793 2500 114 43 49 4-HBB 152 300 2475 4649 1250 114 18 50 6-HBB-102 300 2475 4649 2090 114 28 51 6 HBB 101 170 1886 3235 70 114 0 54 12 HBB-4 2296 4106 360 114 9 58- 8 HBB-6 120 '
3100 10855 13229 2970 200 50 81 3/4-DCB-51 300 2294 3093 1190 114 18 84 3-IICB-19 200 4875 10097 4980 114 74 2 HBB-155 86
- Assume B1 = 1.0 and stress due to weight is 5000 psi k .
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Te: Engineering Report No: GGNS-97.0002. Rev 0
.. Page1of 2 GRAND GUI,F NUCLEAR STATION UNIT 1 SAFETY EVAL,UATION APPLICABILITY REVIEW FORM Document Evaluated: GGNS-97 0002. Rev. O. for Generic Letter 96 06 Evaluation of Drywej A) and Containment Penetrations B) Description of the Proposed Change: The curvose of this Engineerina Report is to communicate the results of the evaluation of Drywell and Containment Penetrations by NPE in accordance with Generic Letter 96-06. The results of the simplified FEA analysis are that all of the penetrations fo.t Grand Gulf meet the required safety fimetiont-ERE-SCREENING Check the applicable boxes below. If any of the following boxes are checked, neither a safety evaluation applicability review nor a safety evaluation is necessary and steps C, D and E may be skipped. The preparer and reviewer must ,ign at the bottom of the form.
The change is editorial as defined on Page 37 of the 10CFR50.59 Safety Evaluation Guidelines.
10CFR$0.59 does not apply per Section 6.1.2.b of PAP 01-S-06-24. Reference An approved safety evaluation covering all tspects of this subject already exists.
Reference SE #
The change, in its entirety, has been approved by the NRC.
[ Reference I' The change is an FSAR change that meets. the exclusion criteria outli:.ed in the 10CFR50.59 Safety Evaluation Guidelines and the change does not represent a change to the Technical 5pecifications or the Technical RequiremensManual.
Safety Evaluation Annlicability Review if any of the following questions in steps C and D are answered "yes", then a full 50.59 Safety Evaluation must be completed.
C) Does the proposed change or activity represent a change to the Technical Specifications?
YES Explain:
NO XX. The ournose of this Engineerina Report is to document the engineerina evaluation for GL 96-06. The results of the heat transfer calculation showed that the anticipated temperature increases were small in relation to the temperature increases expected for eenetration failures. Based on these calculations and engineenng._iudament. none of the oenetrations are exoected to experience temperatures which would lead to a ogetration failure. The oenetrations are not expected to fail durina the boundina accident thus no challenne to containment
N2 l 70): Engitisering Report No: OGNSo97 0002. Rev.0 l Pass 2 of1 inte liv is w;+ Mad. The Fneineerina Raaart documents an evaluation of existinn '
condition and therefore does not alter or imDact any Technical Specifications or Technical Specification Basis.
D)- Does the proposed change or activity represent: i
- (1) A change to the facility which alters, or has the potential to alter, the information, operation, function or ability to perform the fuhetion of a system, structure or component described la the SAR7 YES Explain:
- NO .2RL This Ennineerina Reoort will be uaad to orovide information for reconne to ' '
Ger.eric Letter 96-06. The raanlin of the aimnlified FEA analvals are that all of the r,enetratican for Grand Gulf are operable and that the reauired safety functions are met. The Ensir.e .-inn Reoort daNments an evalaa'lon of existinn condition and therefore does not alter orimoact the SAR.
(2) - A change to a proceduie which alters, or has the potential to alter, a procedure described, outlined or summarized in the SAR7 1 i
- YES Explain: 5 NO .2QL hre is no nic=%e ==rei nvolved w t *i h hia RaMan Raaart (3) A test or experiment not described in the SAR or which requires that a system be operated in an abnormal manner that is not described or previously analyzed in the SAR7 YES Explain:
] There is no test or exoeriment navaciated with this Ennineerina Report. ,
NO .2QL 9
E) is a TRM change requi ? [ ]YES (X]NO PREPARER: J D. Graves v M' Senior Ennineer 1/2
&7 Job Title Date Name / ,
REVIEWER: L La l /z N 4'7 Name" Job Title ' Date ,
if the preparer perfonned an applicability review( screen i ng ), the rev iewer shou ld check below to indicate by which means the independent review reached the same conclusions. >
Reviewed the applicability review documentation.
Completed an independent applicability review.
Performed a verbal review with the preparer, f
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ACCESSION NUMBERS OF OVERSIZE PAGES:
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