ML20029C889
| ML20029C889 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 04/25/1994 |
| From: | Miltenberger S Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20029C890 | List: |
| References | |
| LCR-94-07, LCR-94-7, NLR-N94055, NUDOCS 9405020289 | |
| Download: ML20029C889 (21) | |
Text
(.
t Pubhc Service Electnc and Gas Company St:ven E. Millenberger Pubbe Service Electric and Gas Company P.O. Box 236. Hancocks Bridge, NJ 08038 609-339-4199 v.ce PrhocM an(1 CNef Noctem Oker APR 2 51994 NLR-N94055 LCR 94-07 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
LICENSE AMENDMENT APPLICATION ELIMINATION OF 10CFR50 APPENDIX J, TYPE C TEST REQUIREMENTS FOR LINES PENETRATING CONTAINMENT WHICH TERMINATE BELOW THE TORUS WATER LEVEL HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 This letter submits an application for amendment to Appendix'A of-Facility Operating License NPF-57 for the Hope Creek Generating Station (HCGS), and is being filed in accordance with 10CFR50.90.
Pursuant to the requirements of 10CFR50.91(b) (1), a copy of this request for amendment has been sent to the State of New Jersey.
The proposed amendment involves the elimination of 10CFR50 Appendix J, Type C leak rate testing for certain Containment Isolation Valves, which are located in lines that penetrate the Primary Containment and terminate below the minimum' water _ level in the Suppression Chamber (i.e., Torus).
This request _is based upon an Exemption issued by the NRC on October 30, 1986 for Georgia Power Company's Edwin I. Hatch Nuclear Plant (Docket No.
50-321).
Public Service Electric and Gas Company-(PSE&G) considers 10CFR50 Appendix J, Type C testing, as required by Note 4 of Technical Specification Table 3.6.3-1, to be inappropriate for these valves and proposes the substitution of testing per the applicable requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code.
In response to the NRC Cost Beneficial Licensing' Action (CBLA) initiative, PSE&G met with the NRR Staff on November 12, 1993, to discuss our CBLA Program.
PSE&G considers this submittal a CBLA.
We have estimated the cost savings as $41,600/yr at the HCGS.
Savings over~the life of the plant are $1,331,200.
This proposed change has been evaluated in accordance with 10CFR50. 91 (a) (1), using the criteria in 10CFR50.92(c) and it has been determined that this request involves no significant hazards considerations.
9405020289 940425 O
PDR ADOCK 05000354
.' l p
APR 2 51994 Document Control Desk NLR-N94055 A description of the requested amendment, supporting information and analyses for the change, and the basis for a no significant hazards consideration determination are provided in Attachment 1.
The Technical Specification pages affected by the proposed change are marked in Attachment 2.
Upon NRC approval of this proposed change, PSE&G requests that the amendment be made effective on the date of issuance, but implemented within sixty (60) days to provida sufficient time for associated administrative activities.
Should you have any questions regarding this request, we will be pleased to discuss them with you.
Sincerely, A A A
/
/
~
,/
Affidavit Attachments (2)
C Mr.
T. T. Martin, Administrator - Region I U.
S.
Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. J.
C.
Stone, Licensing Project Manager -
Salem & Hope Creek U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. C. Marschall (SO9)
USNRC Senior Resident Inspector Mr.
K. Tosch, Manager, IV NJ Department of Environmental Protection Division of Environmental Quality i
Bureau of Nuclear Engineering CN 415 i
Trenton, NJ 08625
'I s
REF: NLR-N94055 LCR 94-07 STATE OF NEW JERSEY
)
)
SS.
COUNTY OF-SALEM
)
S. E. Miltenberger, being duly sworn according to law deposes and says:
I am Vice President and Chief Nuclear Officer of Public Service Electric and Gas Company, and as such, I find the matters set forth in the above referenced letter, concerning the Hope Creek Generating Station, are true to the best of my knowledge, information and belief.
.AA i
Subscribed and Sworn o before me thisd day of,
/M/
,1994 n kMI<
1 I f cKew NokarypubliclbfNewJersey KIMBERLY JO BROWN NOTARY PUBL!C Of NEWJERSEY My Commission expires on My comminion Expi,n April 21,1998 l
1 O
1 ATTACHMENT 1 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS i
LICENSE AMENDMENT APPLICATION ELIMINATION OF 10CFR50 APPENDIX J, TYPE C TEST REQUIREMENTS FOR LINES PENETRATING CONTAINMENT 1
WHICH TERMINATE BELOW THE TORUS WATER LEVEL HOPE CREEK GENERATING STATION NLR-N94055 FACILITY OPERATING LICENSE NPF-57 LCR 94-07 DOCKET NO. 50-354 I.
DESCRIPTION OF THE PROPOSED CHANGES As indicated on the marked-up pages in Attachment 2, PSE&G requests that a new Note (i.e., Note 11) be added to supplement Note 4 of Technical Specification Table 3.6.3-1 " Primary
)
. Containment Isolation Valves."
With the addition of this new i
Note, the Technical Specifications will be revised to indicate that 10CFR50 Appendix J, Type C leak testing is not necessary for certain Containment Isolation Valves (CIVs), which serve lines terminating below the minimum water level in the Suppression Chamber (i.e., Torus).
The valves will, however, be tested per the applicable inservice testing requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME B&PV) Code,Section XI - Division 1, Article INV-3000 " Test Requirements," 1983 Edition and Summer 1983 Addenda.
i II. REASON FOR THE CHANGES
\\
The subject license amendment application is a Cost Beneficial j
Licensing Action since it eliminates current 10CFR50 Appendix J, Type C leak rate testing for certain CIVs.
Such testing is not appropriate for these CIVs and is not necessary to ensure that i
post-accident' radiological releases from the Containment are minimized consistent with the existing accident analyses.
Note that this amendment request is not an exemption to 10CFR50 Appendix _J requirements.
10CFR50 Appendix J, Type C testing is not appropriate for certain CIVs on lines which penetrate the Torus and terminate below the minimum water level.
Inservice testing will be performed in accordance with the ASME B&PV Code,Section XI - Division 1 (i.e., ASME XI), Article IWV-3000.
III. JUSTIFICATION FOR TJIE CHANGES q
Primary Reactor Containment leakage test requirements are set j
forth in 10CFR50 Appendix J.
A'CIV is defined in Appendix J, Paragraph II.B 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, Page 1 of 18
ATTACHMENT 1 LCR 94-07 TYPE C TEST ELIMINATION NLR-N94055 Paragraph II.A as
"...an essentially leak-tight barrier.against the uncontrolled release of radioactivity to the environment."
Current interpretation of these definitions is that a CIV is a valve which could represent a potential fission product release pathway to the environment following a postulated accident and consequently, its allowable leakage should be minimized.
Among the CIVs included in Technical Specification Table 3.6.3-1 are those' associated with lines which penetrate the Suppression Chamber and terminate below the minimum water level.
These CIVs are currently subject to Note 4, which states that the valves are leak rate tested with water in accordance with 10CFR50 Appendix J,
Type C requirements.
As discussed in the following paragraphs, the use of the proposed Note 11 will revise the Technical Specifications to clarify that certain CIVs are not Type C tested.
10CFR50 Appendix J, Type C leak rate testing is not appropriate for certain CIVs and is not necessary to ensure that post-accident radiological releases from the Containment are minimized consistent with the existing accident analyses.
Note 11 will also refer to existing ASME XI requirements in the Technical Specifications.
PSE&G committed to perform hydrostatic leak rate testing of the subject isolation valves to determine their leak tightness consistent with Appendix J,Section III.C.3.(b).
This Section provides the acceptance criteria to ensure that "The installed isolation valve seal-water system fluid inventory is sufficient to assure the sealing function for at least 30 days at a pressure of 1.10 Pa."
The purpose of the water leak rate test is to I
ensure a supply of sealing water for thirty (30) days following an accident.
Sealing water, as intended by Appendix J, refers to a protected supply or inventory inside a pipe, which is of sufficient quantity to compensate for post-accident leakage through the CIV(s) serving that line for a 30 day period.
This seal is a barrier against bypass leakage of the post-accident gaseous Primary Reactor Containment atmosphere to the environment.
The CIVs subject to this amendment application are located in piping of systems which penetrate the Torus and terminate below the minimum water level in the Torus.
Since the Torus is designed and operated so that it is filled with water consistent with Technical Specifications 3/4.5.3 " Suppression Chamber,"
3/4.6.2 "Depressurization Systems - Suppression Chamber" and the associated Bases, the supply of water in the Torus is assured during all Design Basis, post-accident modes of operation.
Consequently, the subject isolation valves will remain " sealed" by the water.
The water seal inside the Torus, in conjunction with the design of the piping associated with the penetrations, is a passive, post-accident Containment bypass leakage barrier.
It precludes Page 2 of 18
' ATTACHMENT 1 LCR 94 *'
TYPE C TEST ELIMINATION NLR-N94055 any direct communication between the post-accident Primary Reactor Containment atmosphere and the subject CIVs thereby eliminating the possibility of post-accident Containment bypass leakage.
However, the water seal provided by the Torus is not a
" seal-water fluid system" as intended by Appendix J.
Therefore, 10CFR50 Appendix J, Type C water leak rate testing for the lines and valves is not appropriate and is not necessary to ensure post-accident, containment integrity.
The following paragraphs discuss defense-in-depth considerations which justify the elimination of 10CFR50 Appendix J, Type C leak rate testing and the substitution of ASME XI inservice testing, as appropriate, for the subject'CIVs:
The subject CIVs all serve systems that-are closed systems outside the Primary Reactor Containment.
In addition, the piping inside the Torus is also closed due to its conservative design in conjunction with the water seal provided by the minimum water level in the Torus, as required by the Technical Specifications.
The use of closed systems inside and outside Containment provides passive, redundant isolation barriers to post-accident Containment bypass leakage which will not be adversely affected by single active or passive failures.
Many of the CIVs are designed to remain open during a postulated accident and consequently, post-accident seat leakage is irrelevant for these CIVs.
Other CIVs may be opened or closed post-accident in order to ensure proper system alignment.
These valves will be exposed to water from the Torus on their outboard siden when the systems are operated.
In addition, the adequacy of the extended Containment boundaries following the postulated failure of the CIVs is evaluated in the Hope Creek Generating Station Updated Final Safety Analysis Report (HCGS UFSAR) Section 6.2.4.3.5.
These factors support the conclusion that due to the use of closed piping systems outside the Containment, seat leakage for these valves is not relevant for their function as isolation barriers to post-accident Containment bypass leakage.
The CIVs will, however, remain fully operational and capable of performing all of their required system functions.
Based upon the assessments provided in the attached Table 1, the CIVs will be classified and tested at ASME XI, Category B valves (i.e.,
valves for which seat leakage in the closed position ~is inconsequential for fulfillment of their function).
Check valves will be classified as ASME XI, Category C.
Any leakage through the subject CIVs on lines which terminate below the Torus water level would involve a liquid release and not a gaseous release of the Primary Reactor Containment atmosphere.
The scrubbing effect of the water in the Torus minimizes post-accident fission product releases from the Page 3 of 18
ATTACHMENT l' LCR 94-07 i
TYPE C TEST ELIMINATION NLR-N94055 Containment.
Therefore, a significant radiological consequence would not occur.
The subject CIVs, and associated closed piping systems outside Containment, are all located outside the Primary Reactor Containment in the radiologically-controlled Reactor Building, which is served by the " Filtration, Recirculation, and Ventilation System" (FRVS).
The FRVS is an Engineered Safety Feature and is designed to minimize post-accident radiological releases to the environment.
In addition, the subject systems are included in the Leakage Reduction Program described in Technical Specification 6.8.4.a.
This program is implemented to reduce leakage from systems outside the Primary Containment that could contain highly radioactive fluids following an accident.
Program elements include design features to minimize leakage; instrumentation to detect gross leakage within the Reactor Building; visual examination during system operation; periodic leakage tests; a vigorous corrective action program to correct leakage problems once they have been identified; and preventive maintenance activities.
These elements will detect and correct degradation of the pressure boundaries of the subject systems, thereby reducing post-accident releases and resultant dose consequences in the Reactor Building.
This would also reduce the amount of radioactivity processed and released by the FRVS.
In summary, the lines inside the Torus which terminate below the minimum water level are essentially closed systems inside the Primary Containment.
Consequently, the subject CIVs which serve these lines, cannot communicate directly with the Containment atmosphere.
The use of closed systems outside the Primary Containment ensures that bypass leakage pathways to the environment following a postulated accident do not exist.
Any post-accident leakage (e.g., packing gland leakage) would involve liquid and not gaseous effluent releases.
These releases would' be confined within the Reactor Building and filtered and monitored by the FRVS prior to release from the Plant.
The radiological consequences associated with these releases would not be significant and would be within the existing plant-licensing bases.
Although the subject CIVs are not required to be tested under PSE&G's 10CFR50 Appendix J, Type C testing program, the overall integrated leak-tight integrity of the Primary Reactor Containment will be confirmed during 10CFR50 Appendix J, Type A testing.
The operability of the CIVs will be demonstrated during normal plant operation and through inservice testing, conducted in accordance with the appropriate requirements of ASME XI.
Table 1 identifies the penetrations, associated CIVs, and factors which support the elimination of 10CFR50 Appendix J, Type C testing of the CIVs.
Page 4 of 18 v
' ATTACHMENT 1 LCR 94-07 TYPE C TEST ELIMINATION NLR-N94055 IV.
SIGNIFICANT HAZARDS CONSIDERATION EVALUATION PSE&G has, pursuant to 10CFR50.92, reviewed the proposed amendment to determine whether our request involves a significant hazards consideration.
We have determined that the operation of the Hope Creek Generating Station in accordance with the proposed changes associated with the elimination of 10CFR50 Appendix J, Type C leak rate testing for certain Containment Isolation Valves on lines which penetrate the Torus and terminate below the minimum Torus water level:
1 1.
Will not involve a significant increase in the probability or consequences of an accident previously evaluated.
The Containment Isolation Valves (CIVs) for which Appendix J,
Type C leak rate testing will no longer be performed are all on lines which penetrate the Torus and terminate below the Torus minimum water level.
Since the Torus is designed and operated to be filled with water during and following any postulated Design Basis Accident, the CIVs will remain water sealed during these conditions.
Type C testing of individual CIVs per the requirements of 10CFR50 Appendix J is not necessary since no potential containment bypass leakage path exists due to the water seal and closed system piping.
The CIVs will, however, continue to be tested pursuant to the applicable requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME B&PV)
Code.
In addition, the elimination of 10CFR50 Appendix J, Type C testing for the subject CIVs will not affect the overall leak-tight integrity of the Primary Reactor Containment, which will be demonstrated during 10CFR50 Appendix J Type A Integrated Leak Rate Testing as described in the Hope Creek Generating Station Updated Final Safety Analysis Report (HCGS UFSAR) Section 6.2.6.1.
Consequently, radiological releases and their consequences due.to leakage of the subject CIVs will be minimized and-within the existing plant licensing basis.
2.
Will not create the possibility of a new or different kind of accident from any accident previously evaluated.
This proposal does not involve any hardware or logic changes, nor does it alter the way in which any plant systems operate.
Post-accident Containment isolation features, boundaries and system interfaces are not affected by the changes.
Therefore, there are no new possibilities or types of accidents considered.
Page 5 of 18
ATTACHMENT.1 LCR 94-07 TYPE C TEST ELIMINATION NLR-N94055 i
\\
3.
Will not involve a significant reduction in a margin of safety.
l The proposed elimination of 10CFR50 Appendix J, Type C testing for certain~CIVs in lines which penetrate.the Torus and terminate below the minimum water level will not adversely affect the margins of safety associated with the plant's licensing bases.
The water seal provided by the Torus, in conjunction with the closed system piping, precludes post-accident bypass leakage.
Individual CIVs will be tested in accordance with the ASME B&PV Code Section XI - Division 1, Article IWV-3000, as required.
10CFR50 Appendix J, Type A testing will ensure that the overall Containment leakage rate is consistent with the plant's licensing bases.
In addition, any leakage associated with the subject CIVs will have little radiological significance since it will involve liquid releases (i.e., minimum fission products).
CIV seat leakage will be confined within the closed system piping downstream of the CIVs.
Existing Containment isolation features, boundaries and system interfaces are not affected by the changes.
Since the CIVs and the systems they serve are all located in the Reactor Building, any leakage (e.g., packing gland leakage) which escapes the confines of the closed system piping, will be contained within the Reactor Building.
The Reactor Building is a radiologically controlled area which is served by the Filtration, Recirculation, and Ventilation System.
This assures that all radioactive releases to the environment are within the existing plant licensing bases.
The elimination of Type C testing for_the. subject isolation valves will not affect the existing radiological release evaluations currently described in the HCGS-UFSAR.
The proposed Amendment will not affect the functional l
capability of any plant safety-related structures, systems or components nor will it result in any relaxation of existing plant licensing bases.
In addition, the implementation of the proposed Amendment will result in a reduction of radiological exposure to plant personnel.
Therefore, the proposed revision will-not reduce a margin of safety.
I Page 6 of 18
- i..
. ATTACHMENT 1 LCR 94-07 TEST'C TEST ELIMINATION NLR-N94055
'V.
CONCLUSION Based on the previous discussion, PSE&G has concluded that the proposed change to the Technical Specifications does not involve a significant hazards consideration insofar as the change: (i) does not involve a significant increase in the probability or consequences of an accident previously evaluated, (ii) does not create the possibility of a new or different kind of accident from any accident previously evaluated, and (iii) does not H
involve a significant reduction in a margin of safety.
l 4
Page 7 of 18
TABLE 1 ISOLATION BOUNDARY CONSIDERATIONS SUPPORTING THE ELIMINATION OF 10CFR50 APPENDIX J, TYPE C TESTING FOR CONTAINMENT ISOLATION VALVES ON LINES WHICH PENETRATE THE TORUS AND TERMINATE BELOW THE TORUS WATER LEVEL Penetrations: P-202, HPCI Pump Suction Line P-208, RCIC Pump Suction Line CIVs: BJ-V009 (HV-F042) - HPCI Pump Suction from Torus BD-V003 (HV-F031) - RCIC Pump Suction from Torus Function: These CIVs are located in the piping which supplies water from the Torus 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.
In addition, CIV BD-V003 may be controlled'from the Remote Shutdown Panel.
Since the suction lines are submerged beneath the minimum Torus water level, and the.HPCI and RCIC Systems are closed systems outside 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 Torus, is not an operational concern due to the system configuration, the use of check valves and monitoring instrumentation including Torus water level.
Comments :
1, 3,
4, 5,
and 6 HCGS-UFSAR
References:
HPCI - Fig. 6.3-1 & 7.3-1 sh. 2 of 6 RCIC - Fig. 5.4-8 & 7.4-1 sh. 5 of 5 Designation: Reclassify as ASME XI Category B.
4 mem--mm----m---wwww--mm-mem---
See Comments Key (Page 18) l 1
Page 8 of 18
)
4 TABLE 1 ISOLATION BOUNDARY CONSIDERATIONS SUPPORTING THE ELIMINATION OF 10CFR50 APPENDIX J, TYPE C TESTING FOR CONTAINMENT ISOLATION VALVES ONLLINES WHICH PENETRATE THE TORUS-AND TERMINATE BELOW THE TORUS WATER LEVEL l
l Penetrations: P-203, HPCI Minimum Flow Return Line l
P-209, RCIC Minimum Flow Return Line l
CIVs: BJ-V016 (HV-F012) - HPCI Pump Minimum Flow Bypass to Torus (SV-F019) - RCIC Pump Minimum Flow Bypass to Torus 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 returning flow to the Torus following pump actuation.
Since the HPCI and RCIC Systems are closed systems outside Containment and the minimum flow bypass lines terminate below the minimum water level in the Torus, post-accident bypass leakage is not a concern.
Leakage, in the reverse direction into the Torus, is not an operational concern due to the system configuration, the use of check valves, and monitoring instrumentation including Torus water level.
Comments : 2, 3,
4, 5,
and 6 l
HCGS-UPSAR
References:
HPCI'- Fig. 6.3-1 & 7.3-1 sh. 1 of 6 RCIC - Fig. 5.4-8 & 7.4-1 sh. 5 of 5 i
1
)
Designation: Reclassify as ASME XI Category B.
l i
- - See Comments Key (Page 18)
Page 9 of 18
4 i
l TABLE 1 ISOLATION BOUNDARY CONSIDERATIONS SUPPORTING THE ELIMINATION OF 10CFR50 APPENDIX J, TYPE C TESTING FOR CONTAINMENT ISOLATION VALVES ON LINES WHICH PENETRATE THE TORUS AND TERMINATE BELOW THE TORUS WATER LEVEL
~
Penetrations: P-211A, B,
C, and D, RHR Pump Suction Lines CIVs: BC-V001 (HV-F004 D), BC-V006 (HV-F004B),.BC-V103 (HV-F004A),
~
and BC-V098 (HV-F004C) - RHP. Pump Suction Lines Function: These CIVs are located in the piping providing suction from the Torus to the RHR Purps.
Since the Torus 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.
In l.
addition, CIVs BC-V001 and BC-V006 may be controlled from the Renote Shutdown Panel.
Since the suction lines are submerged beneath the minimum Torus water level and the RHR System is a closed system outside Containment, bypass leakage is not a concern.
Leakage, in the reverse direction into the Torus, is not an operational concern due to the system configuration, the use of pump discharge check valves and valve interlocks, and monitoring instrumentation including Torus water level.
Comments :
1, 3,
4, 5,
6, and 7 HCGS-UFSAR
References:
Fig. 5.4-13 sh. 1 and 2 of 2 Fig. 7.3-7 sh. 4 of 4 Designation: Reclassify as ASME XI Category E.
i
- - See Comments Key (Page 18) i Page 10 of 18
y TABLE 1 ISOLATION-BOUNDARY CONSIDERATIONS SUPPORTING THE ELIMINATION OF 10CFR50 APPENDIX J, TYPE C TESTING FOR CONTAINMENT ISOLATION VALVES ON LINES WHICH PENETRATE THE TORUS AND TERMINATE BELOW THE TORUS WATER LEVEL Penetrations: P-212A and B, RHR System Test, Minimum Flow Bypass, and Jockey Pump Discharge Lines CIVs: BC-V027 (HV-F010B), BC-V028 (HV-
'SB), BC-V125-(HC-F010A)-
and BC-V124 (HV-F024A) - RHR Sy fest Returns to Torus Function: These CIVs are only open to allow full flow testing of their respective RHR trains.
They are normally closed and, if open, will close automatically in response to a Containment Isolation Signal (
i.e.,
High Drywell Pressure and L1 - Low Reactor Water Level).
Since the test return lines are submerged beneath the minimum Torus water level and the RHR System is a closed system outside Containment, p
-accident bypass leakage is not a concern.
Leakage, in the reverse direction into the Torus, is not an operational concern due to the system configuration, the use of check valves and valve interlocks, and monitoring instrumentation including Torus water level.
Comments :
2, 3,
4, 5,
and 6 HCGS-UFSAR
References:
Fig. 5.4-13 sh. 1 and 2 of 2 Fig. 7.3-7 sh. 2 of 4 Designation: Reclassify as ASME XI Category B.
y i
- - See Comments Key (Page 18)
Page 11 of 18
s' TABLE 1 ISOLATION BOUNDARY CONSIDERATIONS SUPPORTING THE ELIMINATION OF'10CFR50 APPENDIX J, TYPE C TESTING FOR CONTAINMENT ISOLATION VALVES ON LINES WHICH PENETRATE THE TORUS AND TERMINATE BELOW THE TORUS WATER LEVEL Penetrations: P-212A and-B, RHR System Test, Minimum Flow Bypass, and Jockey Pump Discharge Lines CIVs: BC-V034 (HV-F007 D), BC-V031. (HV-F007 B), BC-V128 (HV-F007A),
and BC-V131 (HV-F007C) - RHR Pump Minimum Flow Bypass to Torus Function: These CIVs are located on the discharge side of the RHR Pumps and are normally closed.
The CIVs may open after a 10 second time delay when the RHR Pumps are started in order to protect the pumps from possible shutoff head damage.
The CIVs will close automatically in response to RHR' pump. flow conditions in order to satisfy system flow demands.
Accordingly, the CIVs may be open post-accident when returning flow to the Torus following pump actuation.
Since the RHR System is a closed system outside Containment, and the minimum flow bypass lines terminate below the minimum water level'in the Torus, post.
accident bypass leakage is not a concern.
Leakage, in the reverse direction into the Torus, is not an operational concern due to the system configuration, the use of pump discharge check valves, and monitoring instrumentation including Torus water level.
Comments :
2, 3,
4, 5,
and 6 i
.HCGS-UFSAR
References:
Fig. 5.4-13 sh. 1.and 2 of 2 Fig. 7.3-7 sh. 3 of 4 H
Designation: Reclassify as ASME XI Category B.
- -.See Comments Key (Page 18)
Page 12 of 18
TABLE 1 ISOLATION BOITNDARY CONSIDERATIONS SUPPORTING THE ELIMINATION OF 10CFR50 APPENDIX J, TYPE C TESTING FOR CONTAINMENT ISOLATION VALVES ON LINES WHICH PENETRATE THE TORUS AND TERMINATE BELOW THE TORUS WATER LEVEL-Penetrations: P-212A and B, RHR System Test, Minimum Flow Bypass, and Jockey Pump Discharge Lines CIVs: BC-V026 (HV-F011B) and BC-V126 (HV-F011A) - Suppression Pool Return Valves Function: These return valves are permanently out-of-service.
The CIVs are closed during all modes of operation.
Since the return lines terminate beneath the minimum Torus water' level and the RHR System is a closed system outside Containment, post-accident bypass leakage.is not a concern.
Leakage, in the reverse direction into the Torus, is not an J
operational concern due to the system configuration, the use of check valves, and monitoring instrumentation including Torus water level.
Comments :
2, 3,
4, 5,
and 6 HCGS-UFSAR
References:
Fig. 5.4-13 sh. 1 and 2 of 2 Designation: Reclassify as ASME XI Category B.
I e
- - See Comments Key (Page 18) i Page-13 of 18
i l
TABLE 1 ISOLATION BOUNDARY CONSIDERATIONS SUPPORTING THE ELIMINATION OF 10CFRSO APPENDIX J, TYPE'C TESTING FOR CONTAINMENT ISOLATION VALVES ON~ LINES WHICH PENETRATE THE TORUS AND TERMINATE BELOW THE TORUS WATER LEVEL Penetrations: P-212A and B, RHR System Test, Minimum Flow Bypass, and Jockey Pump Discharge Lines CIVs: BC-V260 and BC-V206 - Jockey Pump Discharge Check Valves Function: The check valves are located in the discharge piping downstream of the jockey pumps.
The jockey pumps are normally operating to ensure that the RHR System is full of water and capable of delivering this water as required.
The discharge check valves serve to protect the pumps from damage due to flow reversal as well as to ensure the proper system flow path.
If the jockey pumps are not available post-accident, the discharge check valves close in the absence of jockey pump flow.
Since the RHR Jockey Pump System suction lines and discharge lines are submerged beneath the minimum Torus water level and the RHR System is a closed system outside Containment, post-accident-bypass leakage is not a concern.
Leakage, in the reverse direction into the Torus, is not an operational concern due to the system configuration, the use of check valves, and monitoring _ instrumentation including Torus water level.
Comments : 2, 3,
4, 5,
and 6 HCGS-UFSAR
References:
Fig. 5.4-13 sh. 1 and 2 of 2 Designation: Reclassify as ASME XI Category C.
P i
i
- - See Comments Key (Page 18) i Page 14 of 18
TABLE 1 ISOLATION BOUNDARY CONSIDERATIONS SUPPORTING THE ELIMINATION OF 10CFR50 APPENDIX J, TYPE C TESTING
-FOR CONTAINMENT ISOLATION VALVES ON LINES WHICH PENETRATE THE TORUS.AND TERMINATE-BELOW THE TORUS WATER LEVEL Penetrations: P-216A, B,
C, and D, Core Spray Pump Suction Lines CIVs: BE-V019 (HV-F001B), BE-V020 (HV-F001D), BE-V018 (HV-F001C),
and BE-V017 (HV-F001A) - Core Spray Pump Suction Lines Function: These CIVs are located in the piping which provides suction from the Torus to the Core Spray Pumps.
Since the Torus 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 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 Torus water level and the Core Spray System is a closed systtm outside Containment, post-accident bypass leakage is not a concern.
Leakage, in the reverse direction into the Torus, 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 the suction-side piping
~
of the Core Spray Pumps, and Torus-water level monitoring instrumentation.
Comments :
1, 3,
4, 5,. 6, and 7 HCGS-UFSAR
References:
Fig. 6.3-7 sh. 1 of 1 Fig. 7.3-5 sh. 2 of 2 Designation: Reclassify as ASME XI Category B.
- - See Comments Key (Page 18)
Page 15 of 18
TABLE 1 ISOLATION BOUNDARY CONSIDERATIONS SUPPORTING THE ELIMINATION OF 10CFR50 APPENDIX J, TYPE C TESTING FOR CONTAINMENT ISOLATION VALVES ON LINES j
WHICH PENETRATE THE TORUS AND TERMINATE BELOW THE TORUS WATER LEVEL Penetrations: P-217A and B, Core Spray l Test and Minimum Flow Bypass Lines to Suppression Pool CIVs: BE-V026 (HV-F015B) and BE-V025 (HV-F015A) - Core Spray System Test Returns to-Torus Function: These CIVs are only open to allow full flow testing of their respective Core Spray' trains.
They are normally closed and, if open, will close automatically in response to a Containment Isolation Signal (
i.e.,
High Drywell Pressure and L1 - Low Reactor Water Level).
Since the test return lines are submerged beneath the minimum Torus water level and the core Spray System is a closed system outside Containment, post-accident bypass leakage is not a concern.
Leakage, in the reverse direction into the Torus, is not an operational concern due to the system configuration, the use of check valves, and monitoring instrumentation including' Torus water level.
Comments :
2, 3,
4, 5,
and 6 HCGS-UFSAR
References:
Fig. 6.3-7 sh. 1 of 1 Fig. 7.3-5 sh. 1 of 2 Designation: Reclassify as ASME XI Category B.
P r
- - See Comments Key (Page 18)
Page 16 of 18
TABLE 1 ISOLATION BOUNDARY CONSIDERATIONS SUPPORTING THE ELIMINATION OF-10CFR50 APPENDIX J, TYPE C TESTING FOR CONTAINMENT ISOLATION VALVES ON LINES WHICH PENETRATE THE TORUS AND TERMINATE BELOW THE TORUS WATER LEVEL Penetrations: P-217A and B, Core Spray Test and Minimum Flow Bypass Lines to Suppression Pool CIVs: BE-V036 (HV-F031B) and BE-V035 (HV-F031A) - Core Spray Pump Minimum Flow Bypass to Torus' Function: These CIVs are located on the discharge side of the pumps and are normally closed.
The CIVs may open following a time delay when the Core Spray Pumps are started in order to protect the pumps from possible shutoff head damage.
The valves j
are closed automatically in response to Core Spray Pump flow l
conditions in order to satisfy system flow demands.
Accordingly, the CIVs may open post-accident when returning flow to the Torus
]
following pump actuation.
Since the Core Spray System is a l
closed system outside Containment and the minimum flow bypass lines terminate below the minimum water level in the Torus, post-accident bypass leakage is not a concern.
Leakage, in the reverse direction into the Torus, is not an operational concern due to the system configuration, the use of check valves, and monitoring instrumentation including Torus water level.
Comments :
2, 3,
4, 5,
and 6 HCGS-UFSAR
References:
Fig. 6.3-7 sh. 1 of 1 Fig. 7.3-5 sh. 2 of 2 Designation: Reclassify as ASME XI Category B.
- - See Comments Key (Page 18)
'l
]
Page 17 of 18-1 1
j
4 TABLE 1 ISOLATION BOUNDARY CONSIDERATIONS SUPPORTING THE ELIMINATION OF 10CFR50 APPENDIX J, TYPE C TESTING FOR CONTAINMENT ISOLATION VALVES ON LINES WHICH PENETRATE THE TORUS-AND TERMINATE BELOW THE TORUS-WATER LEVEL COMMENTS KEY 1.
The line is completely-submerged during normal and post-accident operations below the minimum water level in the Torus.
This prevents the Primary Reactor 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 Torus so that it is always submerged post-accident.
This prevents the Primary Reactor Containment atmosphere from impinging on the CIV(s) and precludes bypass leakage post-accident.
3.
The penetration uses a closed system outside the Primary Reactor 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 i
seismic Category I requirements, c) classified as Quality Group B per Regulatory Guide 1.26, and d) is cValuated as an extended Containment boundary following the postulated single failure of the CIV (Refer to HCGS UFSAR Section 6.2.4.3.5).
4.
The penetration arrangement and associated piping.inside the Torus is: a) protected against missiles and pipe whip, b) designed to seismic Category I requirements,.and c)_
classified as Quality Group B per Regulatory Guide 1.26.
The piping terminates below the minimum water level in the Torus and remains water sealed for the duration of the accident even after the assumption of a single active failure.
5.
The piping and CIV(s) in the penetration area are conservatively designed to preclude breach of pressure integrity consistent with HCGS UFSAR Section 3.6.2.1.1.1.
I 6.
The CIV(s) are located in the Reactor Building in an area served by the Filtration, Recirculation, and Ventilation.
System following an accident.
7.
CIV(s) remain open post-accident (i.e.,. seat-leakage.is not relevant in regard to bypass leakage).
Page 18 of 18