ML20135D645
| ML20135D645 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 11/27/1996 |
| From: | Mccoy C GEORGIA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LCV-0909-A, LCV-909-A, TAC-M94881, NUDOCS 9612100072 | |
| Download: ML20135D645 (20) | |
Text
C rgia Pow 3r Company 40 Inepmess Center Parkway J
' Post Office Box 1295 Birmingham, Alabama 35201 Telephone 205 992 7122 m
Georgia Power V P es dent, Nuclear LCV-0909-A November 27, 1996 Docket Nos. 50-424 50-425 TAC-M94881 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen:
VOGTLE ELECTRIC GENEPATING PLANT STAFF REVIEW OF UNRESOLVED INSPECTION ITEM i
50-424/95-27-02 CONCERNING TIIE APPLICATION OF TECIINICAL SPECIFICATIONS TO CONTAINMENT ISOLATION VALVES By letter dated October 4,1996, the NRC staff requested that Georgia Power Company (GPC) provide justification for retaining footnotes d. through h. to FSAR table 16.3-4, or revise the table to eliminate these footnotes. FSAR table 16.3-4 is referenced in the Bases for current Technical Specification (TS) 3.6.3, Containment Isolation Valves, and improved TS 3.6.3, Containment Isolation Valves. This is the list that is used to determine those valves that are subject to the requirements of TS 3.6.3 (both current TS and improved TS). In general, these footnotes state that the requirements of TS 3.6.3 do not apply to these valves, but rather the appropriate system TS applies. A listing of the affected valves and the footnotes applied is provided as Attachment 1.
Valves Associated With the Steam Generator Secondary Side:
The following valves listed on FSAR table 16.3-4 (and shown on Attachment 1) are associated with the steam generator secondary side:
liv-3006A and B; liv-13005A and B; IIV-3016A and B; liv-13007A and B; 11V-3026A and B; IIV-13006A and B; IIV-3036A and B; liv-13008A and B;
')
IIV-7603A; llV-7603B; IIV-7603C; IIV-7603D; 11V-9451; 11V-9452; liv-9453; liv-9454; MMI 11V-5227; IIV-5228; IIV-5229; IIV-5230; IIV-lS196; IIV-15197; IIV-15198; 11V-15199 n40084 9612100072 961127 PDR ADOCK 050004 4 G
U. S'., Nuclear Regulatory Commission
" LCV-0909A Page 2 i
These valves are in lines that penetrate containment and are neither part of the reactor coolant pressure boundary (RCPB) nor connected directly to the atmosphere of the containment. These
)
lines also satisfy the requirements of a closed system. In addition to not being part of the RCPB nor i
connected directly to the atmosphere of the containment, the closed system meets the following additional requirements:
1.
The system is protected against missiles and the effects of high energy line breaks.
i 2.
The system is desi med to Seismic Category 1 requirements.
3.
The system is designed to American Society of Mechanical EngineersSection III, Class 2 requirements.
4.
The system is designed to withstand temperatures at least equal to the containment design temperature.
I 5.
The system is designed to withstand the external pressure from the co- (ainment structural acceptance test.
6.
The system is designed to withstand the design basis accident transient and environment.
For those penetrations associated with the secondary side of the steam generators, the barriers against fission product release to the environment are the steam generator tubes and the piping associated with the steam generators. There is no single active failure that will provide a leakage path for the post loss-of-coolant accident (LOCA) containment atmosphere, and additional post LOCA piping failures are not postulated. The valves associated with these penetrations do not receive a containment isolation signal, do not perform a containment isolation function, and, therefore, are not Type C tested. These valves meet the requirements of 10 CFR 50, Appendix A, Criterion 57 in that they are installed outside of containment in lines that p:netrate primary reactor containment and that are neither part of the RCPB nor connected directly to the containment atmosphere. However, consistent with ANSI 56.8,1994, the steam generator tubes and the piping associated with the steam generators form a boundary that does not constitute a potential primary containment atmosphere pathway during and following a DBA. The requirements of TS 3.6.3 are intended to prevent a direct path between the containment atmosphere and the environs. As i
demonstrated above, the potential for a direct path between the containment atmosphere and the environs does not exist for the subject valves; the valves do not perform a containment isolation function; Type C testing is not applicable to these valves; and, therefore, the requirements of TS l
3.6.3 are not applicable.
l r
U. S'., Nuclear Regulatory Commission
, ' LCV-0909A Page 3 l
Valves Associated With Engineered Safeguards Systems:
The following valves listed on FSAR table 16.3-4 (and shown on Attachment 1) are associated with engineered safeguards systems which are required to function following an accident and are part of a closed system outside containment:
IIV-8801 A and B; llV-8811 A and B; LIV-9002A and B i
Valves HV-8801 A and B, listed on FSAR table 16.3-4 (and shown on Attachment 1), are associated with the centrifugal charging pump cold leg injection path. The centrifugal charging pump cold leg injection path performs an engineered safeguard function following an accident, and it is part of a closed system outside containment. Valves liv-8801 A and B are normally closed valves that are automatically opened on receipt of a safety injection signal and remain open during all modes of emergency core cooling. The centrifugal charging pump cold leg injection path is not a credible leak path for containment atmosphere because the fluid seal within the pipe or the closed piping system outside containment (which will inherently be at a pressure higher than that of containment because of the operating centrifugal charging pump (s)) would preclude release of containment atmosphere to the environs. The source of water is from the refueling water storage tank (RWST) during the injection mode and from the containment emergency sumps during the recirculation mode.
Valves IIV-8811 A and B and HV-9002A and B are for the residual heat removal system (RIIRS) emergency sump suction and the containment spray system (CSS) emergency sump suction, respectively. These valves are normally closed and are opened during the recirculation mode.
When the valves are closed, a water seal is provided by water from the RWST (outside containment during the injection mode) and from the emergency sumps (inside containment) during the recirculation mode. The water seal precludes leakage of the containment atmosphere to the environs.
Furthermore, the TS contain programmatic requirements for minimizing leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. (Current TS 6.7.4.a and improved TS 5.5.2, Primary Coolant Sources Outside Containment.) The charging, residual heat removal, and containment spray systems are included within the scope of these programmatic requirements. By j
minimizing leakage from these systems, the supply of water to maintain the fluid seal is further ensured.
Therefore, valves llV-8801 A and B, IIV-8811 A and B, and IIV-9002A and B are not required to perform a containment isolation function.
, U. S'., Nuclear Regulatory Commission
' LCV-0909A Page 4 Valves Associated With the Nuclear Service Cooling Water System:
The following valves listed on FSAR table 16.3-4 (and shown on Attachment 1) are associated with the nuclear service cooling water system (NSCW):
s liv-2134; HV-2135; HV-2138; HV-2139; HV-1806; IIV-1807; HV-1808; HV-1809; HV-1822; HV-1823; HV-1830; HV-1831 Nuclear service cooling water to and from the containment fan coolers is a closed system inside containment. Therefore, it is not part of the RCPB nor does it communicate with the containment I
atmosphere, and it meets the following additional requirements similar to the steam generator secondary side:
j 1.
The system is protected against missilea and the effects of high energy line breaks.
l l
2.
The system is designed to Seismic Category I requirements.
t 3.
The system is designed to American Society of Mechanical EngineersSection III, Class 2 requirements.
4.
The system is designed to withstand temperatures at least equal to the containment design temperature.
i l
5.
The system is designed to withstand the external pressure from the containment structural acceptance test.
6.
The system is designed to withstand the design basis accident transient and environment.
I l
l Therefore, there is no single active failure that would provide a leakage path for the post LOCA l
containment atmosphere, and additional post LOCA piping failures are not postulated.
In addition, NSCW system operating pressure is higher than the peak LOCA containment pressure.
i Valves HV-2134,2135,2138, and 2139 are required to close on a safety injection signal to isolate J
'aw to nonessential coolers. For these valves, system operating pressure would maintain a fluid seal that would provide additional assurance against leakage of containment atmosphere to the environs. The remaining listed NSCW valves are normally open, and they receive a safety injection signal to open. These valves are required to be open for the containment coolers to perform their '
j safeguard function. For these valves, system operating pressure would also preclude leakage of the j
containment atmosphere to the environs.
4 c
i 1
l U. S. Nuclear Regulatory Commission l
LCV-0909A Page 5 i
These valves meet the requirements of 10 CFR 50, Appendix A, Criterion 57 in that they are installed outside of containment in lines that penetrate primary reactor containment and that are neither part of the RCPB nor connected directly to the containment atmosphere. However, consistent with ANSI 56.8,1994, the containment coolers and the associated piping form a i
boundary that does not constitute a potential primary containment atmosphere pathway during and following a DBA. The requirements of TS 3.6.3 are intended to prevent a direct path between the 1
containment atmosphere and the environs. As demonstrated above, the potential for a direct path between the containment atmosphere and the environs does not exist for the subject valves; the l
valves do not perform a containment isolation function; Type C testing is not applicable to these valves; and, therefore, the requirements of TS 3.6.3 are not applicable.
Exceptions:
l The following valves should not have been listed on FSAR table 16.3-4:
l LV-5242; LV-5243; LV-5244; LV-5245 l
These are the main feedwater regulating bypass valves, and they are upstream (farther away from containment) from the main feedwater isolation valves. The main feedwater isolation valves are the closest valves to containment (other than a check valve), and the main feedwater regulating bypass valves are not qualified as containment isolation valves. Therefore, these valves should not have been added to table 16.3-4, and GPC will initiate action to remove these valves from the table.
1 The following valves may be required to perform a containment isolation function and should not have been noted as exempted from TS 3.6.3:
HV-8105; HV-9001 A and B Valve HV-8105 is a normally open valve in the normal charging line, and it receives a safety
)
injection signal to close to isolate ECCS flow from the normal charging flow path. It is the closest j
l valve to containment, outside of containment, and a check valve is installed in the line inside l
containment. Both the check valve and HV-8105 have been routinely Type C tested in accordance with 10 CFR 50, Appendix J, as set forth in FSAR table 6.2.4-1.
Valves IIV-9001 A and B are in the containment spray discharge lines. These valves are normally closed, and they receive a signal to open upon containment spray actuation. They are located i
outside of containment with a check valve inside containment. Valves HV-9001 A and B and the associated check valves have been routinely Type C tested in accordance with 10 CFR 50, Appendix J, as set forth in FSAR table 6.2.4-1.
l
U. 5., Nuclear Regulatory Commission
' LCV-0909A
[
Page 6 Therefore, GPC will initiate action to revise FSAR table 16.3-4 to remove the footnotes from valves liv-8105, and IIV-9001 A and B that exempt these valves from TS 3.6.3. GPC will apply TS 3.6.3 to these valves in the future and continue to Type C test these valves.
Current Licensing Basis:
The preceding technical basis for our position regarding the subject valves was originally provided to the NRC (in the context of a justification for not performing Type C leakage testing) in response to NRC questions 480.24 and 480.33. (For your ease of review, we have attached to this letter copies of these questions and our responses.) Section 6.2.6 of the VEGP SER (also attached) discusses the NRC staff findings regarding containment leakage testing at VEGP. Specifically, the staff found the leakage testing program to be acceptable and in compliance with GDC 52,53, and 54; Appendix J of 10 CFR 50; and 10 CFR 100. The list of valves (FSAR table 16.3-4) was developed during review of the Unit 1 TS prior to issuance of the low power license. The subject valves were added to the list at the request of the NRC. FSAR table 16.3-4, complete with the footnotes exempting the subject valves from TS 3.6.3 was added to the FSAR in Amendment 32 to the FSAR in December 1986, which was prior to issuance of the low power license for Unit 1 in January 1987. Therefore, it is our position that FSAR table 16.3-4 (with the subject valves and footnotes) is our current licensing basis, and provides adequate protection of the health and safety of the public. Removal of the footnotes would not result in a substantial increase in the overall protection of the public health and safety.
Finally, your request forjustification noted that a search of1 SARs and TSs for other plants licensed around the same time as VEGP Units 1 and 2 did not find a similar situation. GPC has identified similar situations with other licensed facilities, at least with respect to secondary side containment penetrations.
Sincerely, 0h C. K. McCoy CKM/NJS
i U. s. Nuclear Regulatory Commission i
- LCV-0909A Page 7 l
Enclosure xc:
Georgia Power Comnany Mr. J. B. Beasley, Jr.
Mr. M. Sheibani NORMS t
' U. S. Nuclear Regulatory Comnussion i
. Mr. S. D. Ebneter, Regional Administrator Mr. L. L. Wheeler, Licensing Project Manager, NRR Mr.C. R. Ogle, Senior Resident Inspector, Vogtle i
i
)
.. ~.
~
Attichment 1 t
Valves Listed on FSAR Table 16.3 l But Excepted From TS 3.6.3
)
l l
Safety Iniection HV-8811B(d)
RHR emergency sump suction N/A IIV-8811 A(d)
RIIR emergency sump suction N/A f
L IIV-2134(*)
NSCW supply to reactor cavity coolers 40 HV-2138(*)
NSCW return from reactor cavity coolers 40 HV-2135(*)
NSCW supply to reactor cavity coolers 40 HV-2139(*)
NSCW return from reactor cavity coolers
-40 i
HV-8105(d)
Normal charging line 15 HV-1809(*)
NSCW supply to containment coolers N/A
[
t HV-1807(*)
NSCW supply to containment coolers N/A HV-1806(*)
NSCW supply to containment coolers N/A I
r HV-1808(*)
NSCW supply to containment coolers N/A HV-1831(*)
NSCW return from containment coolers N/A
[
HV-1823(*)
NSCW return from containment coolers N/A i
HV-1830(*) '
NSCW return from containment coolers N/A IIV-1822(*)
NSCW return from containment coolers N/A HV-8801A(d)
Boron injection line to cold leg N/A i
HV-8801B(d)
Boron injection line to cold leg N/A Remote Manual i
HV-9002B(0 Containment spray emergency sump suction N/A HV-9002A(0 Containment spray emergency sump suction N/A 1
Attichm:nt i Valves Listed on FSAR Table 16.3-4 l
But Excepted From TS 3.6.3 Containment Spray l
HV-9001A(0 Containment spray supply N/A HV-9001B(0 Containment spray supply N/A 1
Other Automatic Valves i
HV-3006A(s) -
HV-3006B(8)
HV-3016A(8) ~
HV-3016B(8)
HV-13005A(8)
HV-13005B(8)
l HV-13007A(8)
HV-13007B(8)
HV-3026A(8)
. HV-3026B(8)
HV-13008A(8)
HV-13008B(8)
HV-3036A(8)
HV-3036B(8)
HV-13006A(8)
l HV-13006B(8)
i f
HV-7603A*)
Steam generator blowdown 15 2
Att:chment 1 Valves Listed on FSAR Table 16.3-4 But Excepted From TS 3.6.3 Other Automatic Valves (continued)
HV-7603B*)
Steam generator blowdown 15 HV-7603C*)
Steam generator blowdown 15 HV-7603D*)
Steam generator blowdown 15 HV-9451*)
Steam generator secondary side sample 15 HV-9452*)
Steam generator secondary side sample 15 HV-9453*)
Steam generator secondary side sample 15 HV-9454*)
Steam generator secondary side sample 15 HV-5229(8)
HV-5228(8)
HV-5230(8)
HV-5227(8)
HV-15198(8)
HV-15197(8)
HV-15196(8)
l HV-15199(8)
LV-5245(8)
LV-5244(8)
LV-5242(8)
LV-5243(a)
3
Att:chment i Valves Listed on FSAR Table 16.3-4 But Excepted From TS 3.6.3 Footnotes d.
These valves are included for table completeness. The requirements of Technical Specification 3/4.6.3 do not apply; instead the requirements of Technical Specification 4.5.2.e.1 apply. Valve stroke times where specified will be tested pursuant to specification 4.0.5.
These valves are included for table completeness. The requirements of Technical e.
Specification 3/4.6.3 do not apply; instead the requirements of Technical Specification 4.7.4 apply. Valve stroke times where specified will be tested pursuant to specification 4.0.5.
f.
These valves are included for table completeness. The requirements of Technical Specification 3/4.6.3 do not apply; instead the requirements of Technical Specification 4.6.2.1 apply. Valve stroke times where specified will be tested pursuant to specification 4.0.5.
g.
These valves are included for table completeness. The requirements of Technical Specification 3/4.6.3 do not. apply; instead the requirements of Technical Specification 3/4.7.1.5 apply to the main steam isolation and bypass valves and Technical Specifications 3/4.3.2 and 3/4.7.1.6 apply to the main feedwater isolation valves, main feedwater regulating valves, and associated bypass valves, respectively.
h.
These valves are included for table completeness. The requirements of Technical Specification 3/4.6.3 do not apply; instead the requirements of Technical Specification 3/4.7.1.2 apply. Valve stroke times where specified will be tested pursuant to specification 4.0.5.
4 4
UEGP-FSAR-O Question 480.24 Confirm that all fluid lines penetrating containment are listed in table 6.2.4-1, with the isolation valves identified (include test, vent, and drain connections).
Provide justif'ication for each containment isolation valve that will not be Type C (i.e.,
local leak rate) tested.
Response
Table 6.2.4-1 lists all fluid lines which penetrate the containment, and figure 6.2.4-1 shows the corresponding valve arrangements.
For penetrations for which a general design criterion is applicable, only those valves designated as containment isolation valves are listed in table 6.2.4-1.
For fluid penetrations which have no applicable general design criterion (e.g.,
steam generator secondary side), the fluid penetration isolation valves and the isolation valves on the process line that isolate the line from containment atmosphere are identified in table 6.2.4-1.
However, the corresponding valve arrangements in figure 6.2.4-1 show, in addition, the containment penetration test and vent connections (TC and TV, respectively) and other valves in the area of penetration but which are not a part of the isolation provisions.
For more details on the configuration of each penetration, refer to the referenced FSAR figure in table 6.2.4-1 for the 9
respective penetration.
Table 6.2.4-1 identifies those penetrations which will not be Type C tested.
The affected penetrations anc the justification for exemption from Type C testing are discussed below.
A.
Steam generator secondary siae (penetrations 1 through 4,
7 through 10, 11B, 11C, 12B, 12C, 18, 19, 20, 21, and 101 through 104).
These penetrations are associated with the secondary side of the steam generators.
The barriers against fission product release to the environment are the steam generator tubes and the piping associated with the steam generators.
The valves associated with these l
penetrations do not receive a containment isolation l
signal and are not considered containment isolation valves.
Verification of the integrity of this barrier l20 is accomplished in the Type A test.
B Amend. 8 7/84 Amend. 9 8/84 Q480.24-1 Amend. 20 12/85
i VEGP-FSAR-Q l
l B.
Safety injection, residual heat removal, and long-term l
recirculation lines (penetrations 30 through 33, 36 through 39, and 56 through 60).
See the response to vm3+4en 480-3a l
C.
Nuclear service cooling water (penetrations 43 through
)
46, and 91 through 98).
Nuclear service cooling water to and from the containment fan coolers is a closed system inside containment.
The nuclear service cooling water system inside containment is designed and constructed to ASME III, Class 2, and Seismic Category 1 requirements.
No single active failure will provide j
a leakage path for the post-loss-of-coolant accident containment atmosehere.
D.
Reactor coolant pump seal water injection (penetrations 51, 52, 53, and 54).
See the response to question 480.33.
E.
Containment pressure and reactor vessel level instrumentation (penetrations 13C, 14A, 14B, 14C, 67C, EI 69C, 70C, 71C, 85C, 88A, 88B, and 88C).
These 9
penetrations are designed to satisfy the requirements 14 of Regulatory Guide 1.141.
These lines have no isolation valves and rely on a closed system both inside and outside the containment to preclude the release of containment atmosphere to the environment.
The integrity of these penetrations is verified during the Type A tests.
320 F.
Integrated leak rate test connections (penetrations 64A, 64B, 68, and 87).
Following the completion of the integrated leak rate test (i.e.,
Type A test) which is conducted using these penetrations, these penetrations are Type B tested in accordance with 10 CFR 50, Appendix J.
G.
Equipment hatch, personnel locks, and emergency doors.
These penetrations are Type B tested in accordance with 10 CFR 50, Appendix J, except as noted in response t 20 Question 480.34.
H.
Transfer tube (penetration 89).
This penetration is Type B tested in accordance with 10 CFR 50, Appendix J.
l l
l Amend. 9 8/84 Amend. 14 2/85 Amend. 16 4/85 Q480.24-2 Amend. 20 12/85
n a
VEGP-FSAR-Q Question 480.33 FSAR paragraph 6.2.6.3 states that type C testing of the safety injection lines, containment spray lines, and long term recirculation lines will not be done on the basis that these lines are water sealed.
Additional justification is needed for tha elimination of type C tests (note that table 6.2.4-1 indicates type C testing for the spray lines):
A.
For each line, discuss and justify that a sufficient water inventory will be available for at least 30 days following a loss-of-coolant accident (LOCA).
B.
For each line, discuss your plans for hydrostatically testing the valves to show that water leakage from the isolation valves is compatible with the 30-day inventory requirement.
The leakage limits for these valves should be included in the plant Technical Specifications.
C.
FSAR paragraph 6.2.6.3 states that the isolation valves in the charging line of the chemical and volume control system are type C tested using water.
Type C testing using water as the test fluid is permissible only if it can be shown that parts A and B above are satisfied.
Response
The lines penetrating the containment which are required to perform a safeguard function following an accident and are part of a closed system outside containment are not Type C tested.
17 Each of these lines is equipped with isolation valves that can be actuated by the operator from the control room.
Lines which fall into this category are:
A.
Safety injection pump discharge lines (penetrations 30, 9
31, and 33).
B.
Residual heat removal pump discharge lines (penetrations 56, 57, and 58).
l (penetration 32).
leg injection path 17 C.
Centrifugal charging pump cold B
i Amend. 8 7/84 Amend. 9 8/84
'Q480.33-1 Amend. 17 7/85 l
[
VEGP-FSAR-Q D.
Containment emergency sump lines to the residual heat removal and containment spray pumps suction (penetrations 36 through 39).
E.
Residual heat removal pump hot leg suction lines
', penetrations 50 and 60).
Because the safety injection, residual heat removal, long term gi7 sump recirculation, and high head safety injection portions of g
the chemical and volume control system are closed systems outside the containment, designed and constructed to ASME III, Class 2, and Seismic Category 1 requirements, they do not constitute a potential containment atmosphere leak path during or following an accident with a single active failure of a system component.
The above systems are operated and inspected during normal plant operation to ensure that the integrity is maintained.
The isolation valves in the above lines are either normally open at the time of an accident or are opened at some time after the accident to effect immediate and long term core cooling.
Valves which are closed initially or closed at some time following a loss of coolant accident are positioned to effect proper system operation and not to effect a barrier against release of containment atmosphere.
Should the valves leak slightly when closed, the fluid seal within the pipe or the closed piping system outside the containment would preclude release of containment atmosphere to the environs as discussed further below for each penetration.
e Valve HV-8835 (penetration 30) is a power lockout open valve and is closed during hot leg recirculation mode.
When the valve is closed, either one of the safety injection pumps will maintain an inward acting head of water on the valve, except when valves HV-8821A and 17 HV-8821B are closed.
The source of water is the refueling water storage tank (RWST) during the injection mode and the containment emergency sumps during the recirculation mode.
Even if the valve was to leak slightly when closed, and valves HV-8821A and B were closed, the closed system outside containment would preclude release of containment atmosphere to the environs (see figure 6.3.2-1, sheet 3).
o Valves HV-8802A and B (penetrations 31 and 33) are power l
lockout closed valves and are. opened during the long l
term hot leg recirculation mode.
When the valves are closed, either one of the safety injection pumps will maintain an inward acting head of water on the valves (see figure 6.3.2-1, sheet 3).
Amend. 9 8/84 r
Q480.33-2 Amend. 17 7/85 l
l 1
VEGP-FSAR-Q e Valves HV-8801A and B (penetration 32) are normally closed valves that are automatically open on receipt of I
a safety injection signal and remain open during all modes of emergency core cooling.
If one of the valves fails to open, either one of the centrifugal charging
)
pumps will maintain an acting head of water on both sides of the valve.
The source of water is from the RWST during the injection mode and from the containment emergency sumps during the recirculation mode (see figure 6. 3. 2-1, sheet 1).
I Valves HV-8811A and B and HV-9002A and B (penetrations e
36 through 39) are normally closed valves and are opened during the recirculation mode.
When the valves are closed, a water seal is provided by water from the RWST (outside containment during the injection mode) and from the emergency sumps (inside containment) during the recirculation mode.
These valves are enclosed in encapsulation vessels which are an extension of the containment boundary.
The encapsulation vessels are type F ested as discussed in paragraph 6.2.6.2 (see figures 5.4.7-1 and 6.2.2-3),
e Valve HV-8840 (penetration 56) is a power lockout closed 17 valve that is opened during hot leg recirculation mode.
)-
When the valve is closed either one of the residual heat removal (RHR) pumps will maintain an inward acting head of water on the valve except when valve HV-8716A (train A) and/or valve HV-8716B (train B) are closed and the 1
opposite train RHR pump is not running.
The source of water is from the RWST during the injection mode and i
from the containment emergency sumps during the recirculation mode.
Even if the valve was to leak slightly when closed and valve HV-8716A and/or valve HV-8716B are closed and the opposite trair. RHR pump is not running, the closed RHR system outside containment would preclude release of containment atmosphere to the environs (see figure 6.3.2-1, sheet 3 and figure 3
5.4.7-1).
e Valves HV-8809A and B (penetrations 57 and 58) are power lockout opened valves and are closed during hot leg recirculation mode.
When the valves are closed either one of the RHR pumps will maintain an inward acting head of water on the valves except uhen valve HV-8716A and/or s
valve HV-8716B are closed and the same train RHR pump as the closed valve HV-8809A or B is not running.
As discussed above, the closed RHR system outside containment would preclude the release of containment atmosphere to the environs (see figure 6.3.2-1, sheet 3 and figure 5.4.7-1).
Q480.33-3 Amend. 17 7/85
l i
1 VEGF-ESAR-Q l
e Valves HV-8701A and HV-8702A (penetrations 59 and 60) are normally closed and remain closed during the l
injection and recirculation modes.
If these valves were to leak slightly, the RHR pumps would pump the leakage back into the containment and the closed RHR system would preclude any release to the environs (see 'igure 5.4.7-1)
The valve stems are not a potential containment atmosphere leak path when the valves are closed for one of the following reasons:
The valve stem leak-off connection is capped e
(penetrations 30, 31, 32, and 33).
As part of the 1/
inservice inspection program, the system is operated and visually inspected for leakage to ensure that integrity is being maintained.
I e
The valve is enclosed in an encapsulation vessel (penetrations 36, 37, 38, and 39).
)
e The valve stem leakage is routed to the recycle holdup tank which has a diaphragm.
(Penetrations 56, 57, and 58).
The space above the diaphragm is part of the j
negative pressure boundary and is continuously i
ventilated with the safety-related piping penetration exhaust system (see figure 9.3.4.2 and paragraph 9.4.3.2).
The reactor coolant pump seal water injection lines (penetrations 51 through 54) are not required to perform a safety function during an accident.
However, due to the sensitive nature of the seals, it is highly desirable to provide seal flow at all times.
The charging pumps are used for high head safety injection, and flow will be provided by these pumps through the seal injection lines following an accident.
In addition, the system is designed, constructed, and maintained to 9
qualify as a closed system outside the containment.
Each line is equipped with a remote manual containment isolation valve which the operator can close if required.
Table 6.2.4-1 and subsection 6.2.6 have been revised to reflect the above discussion and to indicate that the containment spray pump discharge line valves (penetrations 34 and 35) are type C tested.
l I
t Q480.33-4 Amend. 17 7/85
L..
A purge system is provided for postaccident containment atmosphere cleanup in l
accordance with 10 CFR 50.44.
The system consists of an exhaust penetration line and a filter system.
The containment isolation valves and interconnecting l
piping are seismic Category I.
The system is actuated manually.
The outside isolation valve is locked closed and canually controlled by the operator.
The normally closed remote-manual, parallel isolation valves inside the containment have position indication in the control room.
Th3 post-LOCA reactor cavity hydrogen purge system is designed to preclude
)
hydrogen pocketing following a LOCA by supplying air to the reactor cavity for dilution of the hydrogen that may be released in the reactor cavity area.
The l
syster consists of two redundant 100% capacity fans which automatically start on a safety injection signal.
The fans take suction from the atmosphere within the steam generator compartments and discharge the air into the reactor cavity.
The air flows out of the cavity along the reactor vessel nozzles or through the ventilation openings surrounding the seal ring.
Containment hydrogen mixing is facilitated by the containment fan coolers which take suction from above the operating deck and discharge to the lower levels of the containment. They are, therefore, able to mix the containment atmosphere and prevent hydrogen pocketing in the containment.
The applicant has analyzed the production and accumulation of hydrogen within l
the containment using the guidelines provided in RG 1.7 and SRP Section 6.2.5.
The applicant's analysis snows that a single recombiner, started on the second d:y following the onset of a LOCA, at a containment hydrogen concentration of 3.5 volume percent, is capable of limiting the hydrogen concentration in con-I tainment to below the RG 1.7 lower flammability limit of 4.0 volume percent.
i In FSAR Amendment 14, the applicant confirmed that' emergency operating proce-dures will be in place to ensure that the recombiner system is actuated in a l
l timely manner.
Based on its review of the Vogtle combustible gas control system, the staff concludes that the system satisfies the design and performance requirements of 10 CFR 50.44, the provisions of RG 1.7,
,he requirements of GDC 41, 42, and 43, and the requirements of NUREG-0737, Items II.E.4.1 and II.F.1, Attachment 6, and I
therefore, the system is found acceptable.
6.2.6 Containment Leakage Testing Th2 containment design includes the provisions and features necessary to satisfy l
l the testing requirements of Appendix J of 10 CFR 50.
The design of the contain-ment penetration and isolation valves permits preoperational and periodic leak-age rate testing at the pressure specified in Appendix J of 10 CFR 50.
The staff has reviewed the containment leakage testing program in the FSAR and in response to staff questions, and finds that the proposed reactor containment leakage testing program complies with the requirements of Appendix J of 10 CFR 50 with the exception discussed below.
6-15 Vrgtle SER i
C
The applicant indicated in the FSAR and in response to Q480.33 that lines which penetrate containment and are required to perform a safety function following an accident need not be Type C tested.
The lines that fall into this category include:
(1) the safety injection pump discharge lines (penetrations 30, 31, and 33), (2) the residual heat removal pump discharge lines (penetrations 56, 57, and 58), (3) the centrifugal charging pump cold-leg injection path (pene-tration 32), (4) the containment emergency sump lines to the residual heat re-moval and containment spray pump suction lines (penetrations 36 through 39),
and (5) the residual heat removal pump hot leg suction lines (penetrations 59 and 60).
The applicant has stated that these lines will always be filled th water.
During ECCS operation, water will be flowing into containment, or if the valves are closed the pumps will maintain an inward acting head of water on the valves, assuming a single active failure of a system component.
By letter dated May 21, 1985, the applicant further justified that the valve stems would not become con-tainment atmosphere leak paths for the following reasons.
For penetrations 30 through 33, the valve stem leakoff connections are capped, and surveillance of their condition is included in the inservice inspection program.
For pcnetra-tions 36 through 39, the valves are enclosed in encaps.ulation vessels.
For penetrations 56 through 58, valve sten, leakage is routed to the recycle holdup tank.
For penetrations 59 and 60, the valves are located inside the containment.
Based on the above information, the staff concurs with the applicant that these lines do not constitute potential containment atmosphere leak paths during or following an accioent and, therefore, need not be Type C tested.
i I
With regard to the reactor coolant pump (RCP) seal water injection lines (pene-
_g trations 51 through 54), the applicant states in the response to Q480.33 that p
seal injection is desirable at all times.
Therefore, these lines are not auto-matically isolated in the event of an accident, and are remote manually con-trolled by the operator.
The valves in these penetration lines are designed to
=
9 be open during normal operation as well as postaccident conditions.
In addi-f tion, the system is designed as a safety graded closed system both inside and f
outside containment.
The lines cannot become containment atmosphere leak paths.
Therefore, the staff concludes that Type C testing for penetrations 51 through l
54 is not necessary.
In FSAR Amendment 14, the applicant requested an exemption from certain require-1 ments of 10 CFR 50, Appendix J, Paragraph III.D.2(b)(ii), which states:
" Air i
locks opened during periods when containment integrity is not required by the Io plant's Technical Specifications shall be tested at the end of such period at not less than P,."
i Whenever the plant is in mode 5 fcold shutdown) or mode 6 (refueling), contain-ment integrity is not required.
Hence, if an air lock is opened during mode 5 or mode 6 operations, Paragraph III.D.2(b)(ii) requires that an overall air lock
-)
leakage test at not less than P be conducted prior to entry into mode 4 (hot I
a shutdown).
4 Even if the periodic 6-month test required by Paragraph III.D 2(b)(i) of Appen-dix J has been satisfied, to meet the requirement of Paragraph III.D.2.(b)(ii),
no access to the containment can be allowed while preparing to leave mode 5 un-
[
til an air lock that has been opened in mode 5 or mode 6 is first tested.
The Vogtle SER 6-16 1
test would effectively be required every time mode 5 was entered.
The contain-ment would have to be cleared of personnel while this test was being performed 1
or personnel would have to remain inside containment during the test and until the plant reached mode 4.
Often there are several minor operational and mainte-nance problems that require containment entry just before entering mode 4; the special air lock test would have to wait until all problems requiring containment entry were first corrected.
This is a very restrictive requirement and would i
slow the process of returning to operation.
If the periodic 6-month test of Paragraph III.D.2(b)(i) and the 3-day test re-quirement of Paragraph III.D.2(b)(iii) are current, no maintenance has been per-l formed on the air lock, and the air lock is properly sealed, there should be no reason to expect the air lock to leak excessively just because it has been opened in mode 5 or mode 6.
Accordingly, the staff concludes that the applicant's proposed request for ex-emption from the seal leakage test of Paragraph III.D.2.(b)(ii) is acceptable when the above conditions have been met.
Whenever maintenance has been per-formed on an air lock, the applicant must still meet the requirements of Para-i graph III.D.2(b)(ii) of Appendix J.
I Therefore, an exemption from the requirement of 10 CFR 50, Appendix J, Para-graph III.D.2(b)(ii) is justified and acceptable for Vogtle Units 1 and 2, and appropriate requirements will be added to the plant Technical Specifications.
In view of the staff's findings and conclusions, the staff has determined that, pursuant to 10 CFR 50.12(a), this exemption is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest.
Except as noted, the containment leakage testing program complies with the re-i quirements of Appendix J of 10 CFR 50.
Such compliance provides adequate as-i surance that containment leaktight integrity can be verified throughout service lifetime and that the leakage rates will be periodically checked during service on a timely basis to maintain such leakages within the specified limits of the Technical Specifications.
The plant's Technical Specifications will contain appropriate surveillance requirements for containment leak testing, including test frequencies.
Based on the above review, the staff concludes that the applicant's leakage testing program is acceptable and complies with the require-l ments of GDC 52, 53, and 54; Appendix J of 10 CFR 50; and 10 CFR 100.
6.2.7 Fracture Prevention of Containment Pressure Boundary The staff's safety evaluation review assessed the ferritic materials in the l
Vogtle containment system that constitute the containment pressure boundary to determine if the material fracture toughness is in compliance with the require-ments of GDC 51.
l GDC 51 requires that under operating, maintenance, testing, and postulated acci-dent conditions (1) the ferritic materials of the containment pressure bound-ary behave in a nonbrittle manner and (2) the probability of rapidly propagat-i ing fracture is minimized.
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l Vogtle SER 6-17 iu i