ML20216F755
| ML20216F755 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 09/02/1997 |
| From: | Gordon Peterson DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9709120089 | |
| Download: ML20216F755 (8) | |
Text
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L A t ur I.<1amy Gessuke Newlear Station UV
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- 4800 Gaond Road YrirL SC 2974 Guy R. Nereen
($03) 8314251 omct Yue N ddaet (803) 83IJ426 Mk September 2, 1997 U.S. Nuclear Regulatory Commission Attentions -Document Control Desk Washington, D.C. 20555-0001 Subjects Duke Energy Corporation Catawba Nuclear Station, Unit 1 and Unit 2 Docket No. 50-413, 50-414 Request for Exemption from 10 CFR 50, Appendix A, GDC 57 Catawba Nuclear Station requesta exemption to the General Design Criterion 57 as detailed in 10 CFR Part 50 Appendix A for Main Steam Line Containment Penetrations M261 and M363.
This exemption is specific to two branch lines off of the Main Steam lines that provide steam to the Turbine Driven Auxiliary Feedwater Pump.
These two branch lines each contain a manual gate valve (SA-1 and SA-4, respectively) which is locked in the open position and can only be operated locally.
These valves satisfy two safety functions.
The exemption is needed to provide both the engineered safety functions of providing auxiliary feedwater and containment isolation. The engineered safety function of providing auxiliary feedwater is one of the most risk significant safety functions and is a higher priority, relative to the containment isolation function.
Adding motor operators to valves SA-1 and SA-4 to satisfy containment isolation under all accident conditions would introduce a new failure mode-increasing the probability of an
- accident.
Containment penetrations M261 and M363 are main steam
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penetrations which penetrate primary reactor containment but t
are not part of the reactor coolant pressure boundary or not connected directly to the containment atmosphere.
Outside of containment these main steam lines branch into various separate, individual lines _before reaching-the respective main
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steam isolations valves. -on each of these main steam lines, one branch provides steam-to the turbino driven auxiliary
'feedwater. pump.. Each of these branch lines contains a manual gatefvalve(-SA-1.and SA-4, respectively) which is locked with a' break-away. lock in the open position, to assure'the auxiliary feedwater safety function, and can only be operated locally..
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Document Control Desk Page 2 September 2, 1997 Technical Specification 3.7.1.2.b, Auxiliary Feedwater System, requires that these valves be locked in the open position.
These valves are also listed as Manual Containment Isolation Valves in Technical Specification 3.6.3. Containment isolation for these lines is directed by Emergency and Abnormal procxlures.
There would be no safety enhancement by adding motor operators to those valves.
The basis for this submittal is the identification of an exemption request as an alternative to meeting General Design Criterion 57 in S*fety Evaluation Report for Catawba containment Integrity Positions dated June 27, 1997.
Attached is a detailed technical evaluation and justification for this exemption recuest.
We request review of this matter at your earliest convenience.
Should there be any questions concerning this request, please call Devereux Tower at (803) 831-3419.
Very truly yours('
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b drhd..Peterson Attachment Xct L.A. Reyes, Regional Administrator Region II Senior Resident Inspector Catawba Nuclear Station P.S. Tam, Senior Project Manager ONRR N
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ATTACHMENT Technical Evaluation and Justification a
1 Documpnt Control Desk Attachment Page 1 of 5 September 2,-1997
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Technical Evaluation and Justification for Exemption to GDC 57 Containment Penetrations M261 and M363 are Main Steam (SM) penetrations.
These lines penetrate primary reactor containment and are not part of the reactor cociant pressure boundary or connected directly to the containment atmosphere.
Therefore, these penetrations meet the criteria for GDC 57 applicability.
Outside of containment, these SM lines branch into various separate, individual i
lines before reaching the respective Main Steam Isolation Valve (MSIV).
On each of these SM lines, one such branch forms part of the Main Steam to Aux Equipment (SA) System used to supply steam to drive a Turbine Driven Auxiliary Feedwater (CA) Pump.
The CA Pump Turbine (CAPT), is supplied with steam flow from two separate SM lines via the SA piping.
These two lines are then joined together before reaching the CAPT, such that only one line actually enters the CAPT.
Figure No. 10-6, location H-4 of the Catawba UFSAR shows these valves in the flow diagram for Main Steam to Aux Steam Equipment, diagram No. CN-1593-1.1.
Valves SA-1 (Main Steam B to CAPT Maintenance Isolation) and SA-4 (Main Steam C to CAPT Maintenance Isolation) are manual gate valves located in the Interior Doghouse immediately downstream of the respective SM piping.
SA-1 and SA-4 are currently listed in Catawba Technical Specification (T/S) 3.6.3 (Containment Isolation Valves),
Table 3.6-2 as containment Isolation Valves.
The valves are locked open (with break away locks) and capable of local manual operation only.
These valves are required to be open to supply steam to the CAPT from the respective SM piping for Engineered Safety Features (ESP) operation of the CAPT (Tech Spec 3.7.1.2.b).
However, CAPT operation can continue with one of these valves closed providing that steam is available from the opposite SM piping.
T/S 3.6.3 and Table 3.6-2 are applicable to valves that are required to close on an ESF signal to ensure penetrations passing through containment are isolated during an accident.
Those systems and penetrations essential for reactor core protection are not automatically isolated during such events.
SA-1 and SA-4 are listed as containment Isolation Valves in T/S 3.6.3 Table 3.6-2, but are not closed on an ESP signal.
- Valves SA-3 (S/G B Main Steam to CAPT Stop Check) and SA-6 (S/G C Main Steam to CAPT Stop Check) are stop check valves located in the Aux Building Elevation 543' Mechanical Penetration Room, downstream (CAPT side) of valves SA-1 and SA-4 respectively, before the SA piping joins together to supply the CAPT.
Document Control Desk At tachnient Page 2 of 5 September 2, 1997 The piping and supports for SA-3 and SA-6 are procured and maintained within the constraints of ASME Class II (Duke class B) standardar and are therefore, of acceptable construction to extend the penetration beyond the identified containment Isolation Valves (SA-1 and SA-4).
These valves are required to be open for CAPT operation when supplying steam to the CAPT from the respective SM piping.
However, CAPT operation can continue with one of these valves closed provided steam is available from the opposite SM piping.
Due to accessibility concerns with SA-1 and SA-4 for certain scenarios requiring the isolation of one steam supply to the CAPT, it may be necessary to utilize other valves in these lines for isolation. For the Main Steam Line break, environmental conditions in the Doghouse may prevent Operator access to these valves therefore it would be advantageous to utilize SA-3 and SA-6 for isolation.
The following two accident scenarios require the isolation of one steam supply to the CAPT due to current dose assessment limitations and CA System operation requirements:
1)
Steam Generator Tube Rupture (SGTR) and 2) Main Steam Line Break.
The dose calculations for Main Steam Line Break and SGTR scenarios assume limited release of main steam to the atmosphere.
Operator action is taken as directed within current Emergency Operating Procedures (EP's) and/or Abnormal Operating Procedures (AP's) to manually close SA-1 and/or SA-4 as applicable in these scenarios.
The procedure also provides the option to use SA-3 and SA-6 as required in the event that accessibility to valves SA-1 and SA-4 is limited. SA-1 and SA-4 are Ic Tted in the Interior Doghouse and would not be accessible in the event of a high energy line break (i.e. Main Steam Line Dreak) in this Doghouse.
However, SA-1 and SA-4 are the closest isolation valvet M retainment. If accessible. SA-1 and SA-4 can also be closed in a shorter time frame than SA-3 and SA-6.
During a SGTR, the time required to manually close stop check valves SA-3 and SA-6 rrwy increase due to ' dress-out* requirements and increased radiation monitoring prior to entering the area due to contamination and increased dose in the mechanical penetration room.
Therefore, if accessible, closing SA-1 and SA-4 would be preferable over SA-3 and SA-6.
In each of these accident scenarios, the time required for an Operator to' close the applicable valve has been estimated and factored into the Accident Analyses and resultant dose calculations.
Calculated off-site doses are within allowable values for these scenarios.
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Document Control Desk Attachment Page 3 of 5 September 2, 1997 For the SGTR scenario, failure to isointe steam to the CAPT from the S/G with the tube rupture would allow indefinite release of main steam to the atmosphere via the CAPT exhaust, which would consequently exceed the current dose calculations. Assuming no high energy line break in the Interior Doghouse during a SGTR accident, closing SA-1 and SA-4 would be preferable over SA-3 and SA-6.
For the Main Steam Line Break scencrio, SA-3 or SA-6 (depending on the break scenario) would automatically close (stop check valves) to prevent the diversion of steam from an intact steam line to the faulted piping effectively depressurizing a second S/G, rendering the CAPT inoperable due to the loss of all steam supply, and also affecting the operation / flow balance of the Motor Driven CA punps.
The applicable SA line is isolated as a precaution in case of the check valve fails to close and isolate the faulted steam line.
Unless the steam line break is located in the Interior Doghouse, closing SA-1 and SA-4 would be preferable over SA-3 and SA-6.
Summary of GDC 57 compliance for SA piping section of penetrations M261 end M363:
10CFR50, Appendix A, GDC 57, Reactor containment - Closed System Isolation Valves states that, 'each line that penetrates primary reactor containment and is neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere shall have at least one containment isolation valve which shall be either autonatic, or locked closed, or capable of remote manual operation."
It further states that this valve shall be outside containment and located as close to containment as practical, and that a simple check valve may not be used as the automatic isolation valvo.
Penetrations M261 and H363 do met the requirements of GDC 57 with the exception that valves SA-1 and SA-4 do not automatically close, nor are they locked closed, nor are they capable of remote manual operation.
But valves SA-1 and SA4 serve a dual function, ESF and Containment Isolation. From a Probabilistic Risk Assessment (PRA) perspective, the CAPT is one of the most risk significant safety system components. To enhance the reliability of this ESP system, (CA System) and to help satisfy the requirements of T/S 3.7.1.2.b to have an OPERABLE flow path of the turbine driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply systen valves SA-1 and SA-4 are administrative 1y Locked Open. In emergency situations the Operator's EP (emergency procedures) direct the individuals to close SA-1 and SA-4 at the appropriate times when use l
of steam to the CAPT is no longer necessary or desired. It should be understood that these locks are break away locks and do not require the operator to obtain the key.
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e Document Control Desk Attachrdent Page 4 of 5 September 2, 1997 This exemption request does not change the configuration of the plant as was originally approved in the SER.
It does however add some clarification and rectifies an inconsistency between Technical Specifications and 10 CPR 50 Appendix A GDC 57.
Currently, both Technical Specification and FSAR list SA-1 and SA-4 as Containment Isolation Valves, and the EP guidance is to close SA-1 and SA-4 as applicable.
Manual containment isolation capabilities for the SA piping section of penetrations H261 and M363 cannot be removed from T/S 3.6.3 or from the PSAR as these lines are required to be isolated in certain-scenarios because of current dose assessment limitations and CA Lystem operation requirements.
Although the use of SA-3 and SA-6 for containment isolation has been justified by a previous design study, use of SA-1 and SA-4 as the c:; edited containment Isolation valves better meets the intent of GDC
$7 since they are closer to containment, and based on most of the postulated accidents, SA-1 and SA-4 would probably be more accessible.
To enhance the success of the action to isolate this line for any of the various accident situations the emergency procedures direct the operators to close SA-1 and SA-4 first; however, if unsuccessful the operators would then be directed to close SA-3 and SA-6.
The following applicable Operations EP's and AP's direct Operations to use SA-1 and SA-4 as the first response, with proceduralized option to utilize SA-3 or SA-6 should SA-1 or SA-4 be unavailable for-isolation:
1.
EP/1/A/5000/E-2 (Faulted Steam Generator Isolation) 2.
EP/1/A/5000/E-3 (Steam Generator Tube Rupture) 3.
EP/1/A/5000/ECA-0.0 (Loss of All AC Power) 4.
EP/1/A/5000/ECA-2.1 (Uncontrolled Depressurization of All Steam Generators) 5.
EP/1/A/5000/FR-S.1 (Response to Nuclear Power Generation /ATWS) 6.
EP/1/A/5000/PR-H.3 (Response to Steam Generator High Level) 7.
EP/1/A/5000/FR-P.1 (Response to Imminent Pressurized Thermal Shock) 8.
.EP/1/A/5000/FR-P.2 (Response to Anticipated Pressurized Thermal Shock) 9.
EP/1/A/5000/FR-Z.1 (Response to High Containment Pressure) l
- 10. EP/2/A/5000/E-2 (Faulted Steam Generator Isolation) l w
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bocume:Tt control Desk
- Attachment Page-5 of 5 September.2, 1997' l
11._.EP/2/A/5000/E-3 (Steam Generator Tube Rupture)
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- 12. EP/2/A/5000/ECA-0.0 (Loss _of_All AC Power) i 13.
EP/2/A/5000/ECA-2.1 (Uncontrolled Depressurization of i
All Steam Generators) i 14.'
EP/2/A/9000/FR-S.1
_ Response-to Nuclear Power
(
i Generation /ATWS) 15.-.EP/2/A/5000/FR-H.3 (Response to Steam Generator-High
- Level) 16.
EP/2/A/5000/FR-P.1 (Response to Irreninent Pressurized
. I Thermal Shock) i L
17.
EP/2/A/5000/FR+P.2 (Response to' Anticipated Pressurized Thermal Shock)
- 18.. EP/2/A/5000/FR-Z.1 (Response to High containment
-Pressure) i 19.
AP/1/A/5500/10 (Reactor Coolant Leak) 20.
AP/2/A/5500/10 - (Reactor Coolant Leak) r Adding motor operators to SA-1 and SA-4 would adversely impact the reliability of the CAPT to mitigate an accident because it would j
intre. duce a new failure mode. The assumed time to isolate steam
. gener stor B and C for of fsite dowe consequences using the current plant configuration has already - been found acceptable by _ the NRC during the review of Catawba's Steam Generator Replacement Project dose calculations.
i Therefore, the approval of this request for _ exemption to GDC 57 would not have sny adverse impact on health and safety of the public.
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