ML20086T606
| ML20086T606 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 07/28/1995 |
| From: | Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20086T609 | List: |
| References | |
| NUDOCS 9508030093 | |
| Download: ML20086T606 (16) | |
Text
-
St tion Support Deprrtment
, w 10 CFR 50.90 ,
YECO ENERGY = a */11 % " L e 965 Chesterbrook Boulevard Wayne, PA 19087 5691 July 28,1995 Docket Nos. 50-352 50-353 License Nos. NPF-39 NPF-85 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Subject:
Umerick Generating Station, Units 1 and 2 Technical Specifications Change Request No. 95-13-0 Elimination of Hydrostatic Testing for Certain Valves That Are Assured a Water Seal Following a Design Basis Accident Gentlemen:
PECO Energy Company (PECO) is submitting Technical Specifications (TS) Change Request '
No.95-134, in accordance with 10 CFR 50.90, requesting a change to TS (i.e., Appendix A) of operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2.
The purpose of the proposed TS changes is to eliminate the TS requirement to perform 10 CFR 50, Appendix J, Type C, hydrostatic testing on certain valves that are within closed systems and are assured a water seal following a Design Basis Accident (DBA).
This request is similar to a request made by Georgia Power Company (Edwin 1. Hatch Nuclear Plant, Docket No. 50-321) and approved by the NRC on October 30,1986.
Information supporting this TS Change Request is contained in Attachment 1 to this letter. The proposed changes to the LGS, Units 1 and 2, TS pages are contained in Attachment 2. The TS change information is being submitted under affirmation, and the required affidavit is enclosed.
We request that if approved, the TS Change be issued by January 26,1996, and become effective within 30 days of issuance.
If you have any questions, please do not hesitate to contact us. l Very truly yours, Aa.%
G. A. Hunger, Jr.,
Director-Licensing j
l Enclosure, Attachments cc: T. T. Martin, Administrator, Region 1. USNRC (w/ enclosure, attachments) h yf N. S. Perry, USNRC Senior Resident inspector, LGS (w/ enclosure, attachments) b; R. R. Janati, PA Bureau of Radiation Protection (w/ enclosure, attachments) L 9500030093 950728 '
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f COMMONWEALTH OF PENNSYLVANIA :
- as.
COUNTY OF CHESTER :
W. H. Smith, lil, being first duly swom, deposes and says: That he is Vice President of PECO Energy Company, the Applicant herein; that he has read the enclosed Technical Specifications Change Request No.95-134 " Elimination of Hydrostatic Testing for Certain Valves That Are Assured a Water Seal ,
following a Design Basis Accident," for Limerick Generating Station, Unit I and Unit 2, Facility Operating License Nos. NPF-39 and NPF-85, and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief. ;
v Vice President Subscribed and sworn to before me thisN of 1995.
2- .
hu s i
tentarial Seal Edes 8. Moisenwitz, Putic My ly 10,1999 L
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i ATTACHMENT 1 ,
UMERICK GENERATING STATION UNITS 1 AND 2 Docket Nos.
50452 50-353 Ucense Nos. 1 NPF-39 NPF-85
' Elimination of Hydrostatic Testing for Certain Valves That Are Assured a Water Seal following a Design Basis Accident" ,
Information Supporting Changes - 12 pages 0
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Page 1 of 12 DISCUSSION AND DESCRIPTION OF THE PROPOSED CHANGES PECO Energy Company (PECO Energy) is requesting Technical Specifications (TS) changes which wHI eliminate the requirement to perform 10 CFR 50, Appendix J. Type C hydrostatic tests on certain valves that are assured a water seal following a Design Basis Accident (DBA).
PECO Energy committed to performing hydrostatic testing on High Pressure Coolant injection (HPCI), j Reactor Core Isolation Cooling (RCIC), Core Spray (CS), and Residual Heat Removal (AHR) system ;
containment isolation valves (CIVs) that serve lines below the suppression pool and penetrate below the ,
minimum water level in the suppression pool. Twenty-seven valves per unit are proposed to be changed 1 and are noted as requiring hydrostatic leak testing in TS Table 3.6.3-1 " Primary Containment isolation Valves."
These valves are in portions of the HPCI, RCIC, CS, and RHR systems which were designed and are ,
maintained as closed systems outside of primary containment. The identified valves are assured to remain below the minimum suppression pool water level following a DBA, and the valves are subject to ASME Section XI testing; therefore,10 CFR 50, Appendix J, Type C hydrosiatic tasting of these valves is not necessary to ensure that post accident radiological releases from Primary Conminment are minimized, if approved, TS Table 3.6.3-1, Part A, notation numbers 4,5,19, and 22 for the subject valves, will not be applicable, and will be deleted.
Notation 4 for penetration numbers 203A,B,C.D; 204A,B; 206A,B,C,D; 209; 210; 212; 214; 215; and 226A,B is currently worded as follows: ,
" Inboard gate valve tested in the reverse direction."
Notation 5 for penetration numbers 207A,B; 208B; 216; 235; and 236 is currently worded as follows:
" Inboard globe valve tested in the reverse direction."
Notation 19 for penetration numbers 203A,B,C,D is currently worded as follows:
"The RHR system safety pressure relief valves are flanged to facultate removal and are equipped with double O-ring seal assemblies on the flange closest to primary containment. These seals wHI be leak rate tested by pressurizing between the O-rings, and the results added into the Type C total for this penetration."
Notation 22 for penetration numbers 203A,B,C,D; 204A,B; 206A,B,C,D; 207A,B; 208B; 209; 210; 212; 214; 215; 216; 226A,B; 235; and 236 is currently worded as follows:
" Isolation barrier remains water fuled or a water seal remains in the line post-LOCA. l Isolation valve may be W!ed with water, Isolation valve leakage is not included in 0.60 !
L, total Type B and C tests."
The TS Bases Section 3.6.3, " Primary Containment isolation Valves" is described as follows and will not be changed.
{
'The OPERABluTY of the primary containment isolation valves ensures that the containment ;
atmosphere wHI be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirement of GDC [ General Design Criterion) 54 through 57 of Appendix A of 10 CFR
- 50. Containment isolation within the time limits specified for those isolation valves designed to I
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~' Page 2 of 12 close automatically ensures that the release of radioactive material to the environment wNI be consistent with the assumptions used in the analyses for a LOCA."
If approved, the following new notation wNI be added to TS Table 3.6.3-1, referenced by the subject valves. ;
"These valves are in lines that are below the minimum water level in the wppression pool, are part of closed systems outside primary containment, and are in portions of lines which a water seal wHl be present following an accident. Therefore,10CFR50, Appendix J, Type C testing is not required "
, TS Umiting Condition For Operation Section 3.6.1.2.d, requiring hydrostatic testing, TS Surveillance Requirement (SR) 4.6.1.2.d.3, noting approval of exemptions permitting hydrostatic testing, and TS SR 4.6.1.2.g, required frequency for hydrostatic testing, wHl not be changed since there are other valves that will continue to require hydrostatic testing.
TS SAFETY ASSESSMENT Currently, PECO Energy is committed, as described in the LGS NRC SER (NUREG - 0991), dated August 1963, Section 6.2.6.4, to perform hydrostatic testing on HPCI, RCIC, CS, and RHR valves that serve lines below the suppression pool and penetrate below the minimum water level in the suppression pool. Twenty-seven valves are included in the scope of this request and are described in Exhibit 1, which is attached to this Safety Assessment.
The NRC SER states, " Appendix J to 10 CFR 50 requires that unless valves are pressurized with fluid from a seal system, they shall be pressurized with air or nitrogen for leak testing purposes (Paragraph Ill.C.2). There are a number of liquid fuled systems, however, that are specifically designed to remain intact after a LOCA and thus provide a water seal for the system isolation valves or ensure that only >
liquid leakage from the containment wNI occur, Because of this, the applicant [PECO) proposes to perform hydrostatic testing to determine the leak tightness of the subject valves. The hydrostatic valve leakages are excluded from the combined leakage rate for all penetrations and valves as provided for in Appendix J, Paragraph Ill.C.3."
The piping for these systems which includes the subject valves penetrates the suppression pool and terminates below the minimum water level of the suppression pool. The suppression pool water level is assured under post-accident conditions, per the LGS Updated Final Safety Analysis Report (UFSAR) i Section 6.2.3.2.3.1, whereby these valves wHl remain sealed with water 30 days following the postulated accident. The suppression pool level is designed and operated so that water level is maintained in accordance with TS 3/4.5.3, " Suppression Chamber," 3/4.6.2, "Depressurization Systems - Suppression Chamber," and the associated TS Bases. LGS calculation MISC-62 determined that the lowest water level that the suppression pool will experience after a DBA is at least 4 feet above affected penetrations.
The supply of water in the suppression pool is assured for 30 days during all DBA, post-accident modes of operation. Since the suppression pool is analyzed to remain fHied with water at a minimum level above the penetrations, no leak test is necessary to satisfy Appendix J requirements, and these lines should not be considered as Appendix J penetrations.
The affected valves may be open, or change state, post-accident to support the design function of their associated Emergency Core Cooling System (ECCS). Therefore, the reliance to limit containment bypass leakage is placed on a suppression pool water seal and the integrity of the closed systems. LGS TS Section 6.8.4.a " Primary Coolant Sources Outside Containment" establishes a program to monitor and control leakage from systems located outside containment that could contain radioactive fluida during a serious transient or accident. This program applies to the ECCS affected by these proposed changes, and ensures that leakage into secondary containment (e.g., packing, flanges, seals) is controlled. Previous leakages from these systems have been found to be very low and well below the i established limit for these systems. This program wNi ensure leakage is limited and maintained below y
, . _ . y
- Page 3 of 12 established limits which were previously considered in the LGS Safety Analysis Report (SAR). The proposed change wRI not contribute to higher levels of system leakage, and any leakages from these systems are processed via standby gas treatment and the radwaste system in order to maintain ALARA and comply with regulatory guidance.
10 CFR 50, Appendix J, Paragraph II.B. defines a Containment Isolation Valve (CIV) as "any valve which is relied upon to perform a containment isolation function". The term " containment" refers to primary reactor containment, which is defined in Appendix J. Paragraph II.A as "the structure or vessel that encloses the components of the reactor coolant pressure boundary, as defined in 10 CFR 50.2(v), and serves as an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment". Based on these definitions, an Appendix J, CIV is a valve which could represent a potential fission product release pathway from the containment atmosphere to the environment following a postulated accident, and therefore its allowable leakage limit should be minimized.
10 CFR 50, Appendix A, V " Reactor Containment," General Design Criterion (GDC) 50 states "The containment structure, including access openings, penetrations, and the containment heat removal system, is designed so that the containment structure and its intemal compartments can accommodate, without exceeding the design leakage rate [1/2 % by weight of containment air per day) with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident." l The potential leakages through the affected penetration valves would be liquid leakages, and the leakage would be into a closed piping system outside the containment; thus these leakages are not part of the permissible 1/2 % per day air leakage.
These valves are contained in lines that penetrate the suppression pool and that are assured to remain water sealed for 30 days after the onset of an accident; therefore, they wHI not be exposed to primary containment atmosphere.
In addition to the above definitions, ANSI /ANS 56.8-1994 Section 3.3.1 states " Primary containment
! boundaries not requiring Type B or Type C testing include: (1) boundaries that do not constitute potential primary containment atmospheric pathways during and following a DBA..."
The affected valves WHI continue to be tested per the applicable inservice testing (IST) requirements in accordance with ASME Section XI, under the LGS IST Program. The CIVs will be reclassified from ASME Section XI, Category A valves to Category B valves. Category B valves are those for which seat Icakage in the closed position is inconsequential for fulfillment of the required function (s). A LGS IST Program change wHi be necessary in order to reclassify the valves and will be performed under the provisions of 10 CFR 50.59 following NRC approval of this TS change.
Finally, the affected penetrations wHI continue to be subjected to the periodic 10 CFR 50, Appendix J.
Type A test (Integrated Containment Leakage Rate Test).
Information Sunoortina a Findina of No Slanificant Ha7ards Consideration We have concluded that the proposed changes to the Limerick Generating Station (LGS) Unit 1 and Unit 2 TS to eliminate the requirement to perform 10 CFR 50, Appendix J, Type C hydrostatic testing on certain valves that are within closed systems and are assured a water seal following a Design Basis Accident (DBA), do not involve a Significant Hazards Consideration. In support of this determination, an evaluation of each of the three (3) standards set forth in 10 CFR 50.92 is ptovided below.
- 1. The oroonaad TS chanaes do not involve a slanificant increase in the probabHity or conseauences of an accident oreviousiv evaluated.
The pjntsy containment (drywell and suppression pool) and the affected closed systems are accident mitigators not accident initiators. The proposed change to the scope of Appendix J, Type C testing does not affect the probabHity of the DBA. The
Page 4 of 12 valves wuI continue to be maintained in an operable state, and in their current design configuration. There is no correlation between the scope of Appendix J. Type C testirq and accident probabNity. There are no physical or operational changes to the containment structure, system or components being made as a resu:t of the proposed changes. Therefore, the consequences of a malfunction of equipment important to safety is not increased from those previously evaluated.
The consequences of loss-of-coolant accidents (LOCAs) under the proposed change were considered where a single active faHure of a containment isolation valve (ClV) or a passive faNure of the closed system were reviewed, within the limits of the existing licensing basis. Under the existing licensing basis, a pipe rupture of the seismically quallflod ECCS piping does not have to be assumed concurrent with the LOCA, except if it !s a consequence of the LOCA. Consideration of consequential faHures can be eliminated, since a LOCA inside containment is separated from the affected piping by the containment structure. Consideration of consequential fauures of the ECCS piping from LOCAs outside containment are outside the Appendix J design considerations. A single active faHure of the CIV, under the LOCA condition, can be accommodated since the closed and water sealed system piping remains as the leakage barrier. The ECCS passive faHure criterion does require consideration of system leaks, but not pipe breaks, beyond the initiating LOCA. The capability to make-up water inventory to the suppression pool is adequate to ensure that postulated seat leakage and pipe leakage does not result in a condition that jeopardizes pool level. Make-up capabuity exists for the suppression pool via the Condensate Storage Tank and Ultimate Heat Sink Spray Pond. Operator actions to make-up the suppression pool are delineated in ex: sting !
Operating Procedures.
The subject valves are single isolation valves associated with ilnes that penetrate the l primary containment, but are not connected directly to the primary containment atmosphere or the reactor coolant pressure boundary. This configuration is described in the LGS UFSAR, Section 6.2.4.3.1.3.1, which states "the systems which the lines from the suppression pool connect to outside containment are closed systems meeting the appropriate requirements of closed systems." The integrity of these closed systems are also monitored and controlled in accordance with TS Section 6.8.4.a. Any leakage that may escape the confines of the closed system wWI be contained within the Reactor Building, treated by standby gas and radwaste systems, and, therefore, are within the existing LGS licensing bases.
Finally, the affected penetrations will continue to be subjected to the periodic 10 CFR 50, Appendix J. Type A test (Integrated Containment Leakage Rate Test).
The suppression pool level is designed and operated so that water lovel is maintained in accordance with current TS, and the associated bases. The supply of water in the suppression pool is assured for 30 days during all DBA, pos'-accident modes of operation. The lowest water level which the suppression pool will reach was analyzed, and it was determined that the affected lines wNI remain below this minimum level, thereby assuring a water seal. The valves wHI continue to be tested and maintained to ensure their operabHity, and the closed systems' integrity wHI continue to be monitored and controlled in accordance with TS 6.8.4.a and the performance of the periodic 10 CFR 50, Appendix J, Type A test.
Therefore, the proposed changes wHI not increase the probabHity or consequences of an accident previously evaluated.
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Page 5 of 12
- 2. The orW TS chances do not craa*= the _e-WV of a new or different kind of Edent from any accident previously evaluated.
The proposed changes do not change the plant response to accident scenarios, and do not introduce nevi or different scenarios. The primary containment (drywell and suppression pool) and the affected closed systems are accident mitigators not accident initiators. 'I?.e propM change to the scope of Appendix J, Type C hydrostatic testing maintains the twisting bairiers to primary containment bypass leakage by the assurance that a water seal will be maintained for 30 days during all DBA, post-ace'Jent modes of operation. The valves will continue to be tested and maintained to ensure their operability, and the closed systems' integrity will continue to be monitored and controC is accordance with TS 6.8.4.a. Therefore, the proposed changes cannot cause an accident, and the plant response to the design basis events is unchanged, whereby the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. The orooosed TS chances do not InvcAve a sionificant reduction in a maroin of safety.
The water seal provided by the assurance of a minimum suppression pool level will prevent post accident containment bypass leakage. Appendix J does not require air leak testing of the valves since the 30 day post-accident supply of water is maintained.
in addition, the closed systems' integrity is monitored and controlled in accordance with TS 6.8.4.a. Any leakage that may escape the confines of the closed system will be contained within the Reactor Building, and is within the existing LGS licensing bases.
Therefore, the proposed TS changes do not involve a significant reduction in a margin of safety. l INFORMATION SUPPORTING AN ENVIRONMENTAL ASSESSMENT An Environmental Assessment is not required for the Technical Specifications changes proposed by this request because the requested changes to the LGS, Units 1 and 2, TS conform to the criteria for
" actions eligible for categorical exclusion," as specified in 10 CFR 51.22(c)(9). The requested TS changes will have no impact on the environment. The proposed TS changes do not involve a Significant Hazards Consideration as discussed in the preceding Safety Assessment section. The proposed changes do not involve a significant change in the types or significant increase in the amounts of any effluent that may be released offsite. In addition, the proposed TS changes do not invche a significant increase in individual or cumulative occupational radiation exposure.
CONCLUSION The Plant Operations Review Committee and the Nuclear Review Board have reviewed these proposed changes to the LGS Unit 1 and 2 TS, and have concluded that the changes do not involve an unreviewed safety question, and will not endanger the health and safety of the public.
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a Page 6 of 12 Exhibit 1 List of Containment isolation Valves (ClVs) on Lines Which Penetrate the Suppression Pool and Terminate Below the Minimum Suppression Pool Water Level Penetrations: X-209, HPCI Pump Suction Lin0 X-214. RCIC Pump Suction Une CIVs: HV1(2)55F042 - HPCI Pump Suction from Suppression Pool HV1(2)49F031 - RCIC Pump Suction from Suppression Pool Function: These CIVs are located in the pip!ng which supplies water from the Suppression Pool to the HPCI and RCIC pump suction nozzles. Since the Condensate Storage Tank (CST) is the preferred source of water to the HPCI and RCIC pumps, these CIVs are normally closed. They may open post-accident either automatically in response to low CST water level or remote-manually from the Control Room. The HPCI Suction from the Suppression Pool may auto-open on high Suppression Pool level. Since the sucticn lines are submerged beneath the minimum Suppression Pool water level, and the HPCI and RCIC Systems are closed systems outside Primary Containment, post-accident bypass leakage either with the HPCI and RCIC pumps operating, or removed from service, is not a concern.
Leakage in the reverse direction into the Suppression Pool is not an operational concern due to the system configuration, the use of check valves, and monitoring instrumentation including Suppression Pool water level.
Comments, See Key: 1, 3, 4, 5, and 6 LGS-UFSAR
References:
HPCI - Fig. 6.2-36 detail 42 ,
RCIC - Fig. 6.2-36 detail 42 j
Page 7 of 12 Perietrations: X-236, HPCI Pump Recirculation Line X-216, RCIC Pump Recirculation Line CIVs: HVI(2)55F012 - HPCI Pump Minimum Flow Bypass to Suppression Pool i HVI(2)49F019 - RCIC Pump Minimum Flow Bypass to Suppression Pool
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Function: These CIVs are located on the discharge side of the HPCI and RCIC Pumps.
The valves are normally closed and operate automatically in response to HPCI and RCIC pump flow conditions in order to protect the pumps from possible shutoff head damage while satisfying flow demands. Accordingly, the CIVs may open post-accident when retuming flow to the Suppression Pool following pump actuation. Since the HPCI and RCIC Systems are closed systems outside Primary Containment and the minimum flow bypass lines terminate below the minimum water level in the Suppression Pool, post-accident bypass leakage is not a concem.
1 Leaka0e into the Suppression Pool is not an operational concem due to the system configuration, the use of check valves, and monitoring instrumentation including Suppression Pool water level.
Comments, See Key: 2, 3, 4, 5, and 6 LGS-UFSAR
References:
HPCI - Fig. 6.2-36 detail 24 RCIC - Fig. 6.2-36 detail 24 Penetrations: X-203A,B,C,D, RHR Pump Suction Lines CIVs: HV1(2)51F004A,B,C,0 - RHR Pump Suction Unes PSV1(2)51F030A,B,C,D - RHR Pump Suction Relief Valves Function: These CIVs are located in the piping providing suction from the Suppression Pool to the RHR Pumps. Since the Suppression Pool is the source of water for the RHR Pumps, these CIVs are maintained open during normal and post-accident operation. This assures a supply of water is available to support RHR System operation. They may only be closed remote-manually from the Control Room. Since the suction lines are submerged beneath the minimum Suppression Pool water level and the RHR System is a closed system outside Primary Containment, bypass leakage is not a concern.
Leakage in the reverse direction into the Suppression Pool is not an operational concem due to the system configuration, the use of pump discharge check valves and valve interlocks, and monitoring instrumentation including Suppression Pool water level.
Comments, See Key: 1, 3, 4, 5, 6, and 7 LGS-UFSAR
References:
Fig. 6.2-36 detail 16 i
Page 8 of 12 Penetrations: X-226A,B, RHR Recirculation Line CIVs: HVI(2)51-105A,B RHR Pump Recirculation Flow to Suppression Pool Function: These CIVs are located on the discharge side of the RHR Pumcw and are normally open. Tae valves may only be closed remote manually from the Control Room. Accordingly, the CIVs may be open post-accident when retuming flow to the Suppression Pool following pump actuation. Since the ;
RHR System is a closed system outside Primary Containment, and the minimum flow bypass lines terminate below the minimum water level in the Suppression Pool, post accident bypass leakage is not a concem Leakage !nto the Suppression Pool is not an operational concem due to the system contiguration, the use of pump discharge check valves, and monitoring ,
instrumentation including Suppression Pool water level. ,
Comments, See Key: 2, 3, 4, 5, 6, and 7 LGS-UFSAR
References:
Fig. 6.2-36 detail 35 Penetrations: X 206A,B,C,D, Core Spray Pump Suction Lines CIVs: HV1(2)52F001 A,B,C,D - Core Spray Pump Suction Lines !
Function: These CIVs are located in the piping which provides suction from the Suppression Pool to the Core Spray Pumps. Since the Suppression Pool is the source of water for the Core Spray Pumps, these CIVs are maintained open during normal and post-accident operation. This assures a supply of water is 1 available to support Core Spray System operation. They may only be closed remote-manually from the Control Room. Since the suction lines are submerged beneath the minimum Suppression Pool water level and the Core Spray System )
is a closed system outside Primary Containment, post-accident bypass leakage is not a concem.
Leakage in the reverse direction into the Suppression Pool is not an operational concern due to the system configuration, the use of pump discharge check valves, procedural controls used when aligning the Condensate Storage Tank to l the suction-side piping of the Core Spray Pumps, and Suppression Pool water level monitoring instrumentation.
Comments, See Key: 1, 3, 4, 5, 6, and 7 LGS-UFSAR
References:
Fig. 6.2-36 detail 42
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Page 9 of 12 Perietrations: X-207A,B, Core Spray Test Line CIVs: HVI(2)52F015A,B Core Spray System Test Returns to Suppression Pool Function: These CIVs are only open to aCow full flow testing of their respective Cort, Spray trains. They are normally closed and, if open, will close automatically in response to a Core Spray Initiation Signal (i.e., High Drywell Pressure and Low Reactor Water Level) Sirice the test return lines are submerged beneath the minimum Suppression Pool water level and the Core Spray System is a closed system outside Primary Containment, post accident bypass leakage is not a concern.
Leakage into the Suppression Pool is not an operational concem due to the system configuration, the use of check valves, and monitoring instrumentation including Suppression Pool water level.
Comments, See Key: 2, 3, 4, 5, and 6 LGS-UFSAR
References:
Fig. 6.2-36 detail 24 Penetrations: X-2088, X-235, Core Spray Pump Recirculation Line HVI(2)52F031 A,B - Core Spray Pump Minimum Flow Bypass to Suppression Pool !
CIVs:
Function: These CIVs are located on the discharge side of the pumps and are normally open. The valves are closed automatically in response to Core Spray Pump flow conditions in order to satisfy system flow demands. Accordingly, the CIVs may open post-accident when returning flow to the Suppression Pool following pump actuation. Since the Core Spray System is a closed system outside Primary Containment and the minimum flow bypass lines terminate below the j minimum water level in the Suppression Pool, post-accident bypass leakage is !
not a concern.
Leakage into the Suppression Pool is not an operational concern due to the l system configuration, the use of check valves, and monitoring instrumentation including Suppression Pool water level.
Comments, See Key: 2, 3, 4, 5, and 6 LGS-UFSAR
References:
Fig. 6.2-36 detail 24 I
Page 10 of 12 Perietrations: X-204A,B, RHR Pump Test Line CIVs: HVI(2)51 125A,B - RHR Pump Test une Function: These CIVs are maintained open during normal and post-accident operation.
They allow for periodic testing of the performance characteristics of the RHR pumps and to demonstrate the capabNity of the system. The lines provide a return flow path for suppression pool cooling mode and serve as pumps A & B min flow lines. The valves may only be closed remote-manually from the Control Room. Since the test lines are subirmiged beneath the minimum Suppression Pool water level and the RHR system is a closed system outside Primary Containment, bypass leakage is not a concern.
Comments, See Key: 2, 3, 4, 5, 6, and 7 LGS-UFSAR
References:
Fig. 6.2-36 detaR 36 Penetrations: X-215, RCIC Turbine Exhaust Line CIVs: HV1(2)49F060 RCIC Turbine Exhaust to Suppression Pool i
Function: These CIVs are located in the RCIC Turbine Exhaust piping to the Suppression Pool. These CIVs are maintained open during normal and post-accident operation. This assures that RCIC turbine steam can be exhausted into the Suppression Pool through the spargers when the RCIC pump has to be operated. The CIVs may only be closed remote-manually from the Control ,
Room. Since the exhaust line is submerged beneath the minimum Suppression Pool water level and the system is a closed system outside Primary Containment, bypass leakage is not a coricern.
Comments, See Key: 2, 3, 4, 5, 6, and 7 LGS-UFSAR
References:
Fig. 6.2-36 detail 24
Page 11 of 12 Penetrations: X-212, HPCI Pump Test and Flush Une CIVs: HV1(2)55F071 - HPCI System Test Retum to Suppression Pool Function: These CIVs provide Isolation for the flushing path to the Suppression Pool. They are normally closed, and if open, will close automatically in response to a Primary Containment Isolation Signal (i.e., High Drywell Pressure and Low ;
Reactor Water Level). Since the test retum lines are submerged beneath the ,
minimum Suppression Pool water level and the HPCI system is a closed system I outside Primary Containment, post-accident bypass leakage is not a concem. )
i Leakage into the Suppression Pool is not an operational concern due to the system configuration, and monitoring instrumentation including Suppression Pool water level. j Comments, See Key: 2, 3, 4, 5, and 6 LGS-UFSAR
References:
Fig. 6.2-36 detail 24 Penetrations: X-210, HPCI Turbine Exhaust Line CIVs: HV1(2)55F072 - HPCI Turbine Exhaust to Suppression Pool Function: These CIVs are located in the HPCI Turbine Exhaust piping to the Suppression Pool. These CIVs are maintained open during normal and post-accident ;
operations. The CIVs may only be closed remote-manually from the Control Room. Since the exhaust lines are submerged beneath the minimum Suppression Pool water level and the HPCI system is a closed system outside Primary Containment, bypass leakage is not a concern.
Comments, See Key: 2, 3, 4, 5, 6, and 7 LGS-UFSAR
References:
Fig. 6.2-36 detall 24 l
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f Page 12 of 12 COMMENTS KEY
- 1. The line is completely submerged during normal and post-accident operations below the minimum water level in the Suppression Pool. This prevents the Primary Containment atmosphere from impinging on the CIV(s) and precludes bypass leakage post-accident.
- 2. The line terminates below the minimum water level in the Suppression Pool so that it is always submerged post-accident. This prevents the Primary Containment atmosphere from impinging on the CIV(s) and precludes bypass leakage post-accident.
- 3. The penetration uses a closed system outside the Primary Containment as a second isolation barrier in addition to the identified CIV(s). The system is: a) protected against missiles and pipe whip, b) designed to seismic Category I requirements, c) classified as ASME Section lli class 2, and d) designed to temperature and pressure conditions that the system will encounter post-LOCA. The system is included in the scope of the 10 CFR 50, Appendix J, Type 'A' test.
- 4. The penetration arrangement and associated piping inside the Suppression Pool is: a) protected against missiles and pipe whip, b) designed to seismic Category I requirements, and c) classified as ASME Section ill, class 2. The piping terminates below the minimum water level in the Suppression Pool and remains water sealed for the duration of the accident.
- 5. The piping and CIV(s) in the penetration area are conservatively designed to preclude breach of pressure integrity consistent with LGS UFSAR Section 6.2.1.1.1.
- 6. The CIV(s) are located in the Reactor Building in an area serviced by the Standby Gas Treatment System (SGTS) following an accident. ;
- 7. CIV(s) remain open post-accident and seat leakage is not relevant in regard to bypass leakages.
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ATTACHMENT 2 UMERICK GENERATING STATION UNITS 1 AND 2 ;
Docket Nos. i 50-352 50453 License Nos.
NPF 39 NPF-85 t TECHNICAL SPECIFICATIONS CHANGE REQUEST NO. 95134 4
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UNIT 1 UNIT 2 3/4 6-26 3/4 6-26 3/4 6-27 3/4 6-27 3/4 6-29 3/4 6-29 3/4 6-43 3/4 6-43 I
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