ML20206Q357
| ML20206Q357 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 01/06/1999 |
| From: | CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20206Q354 | List: |
| References | |
| NUDOCS 9901140172 | |
| Download: ML20206Q357 (96) | |
Text
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Attichment 3 l py.CEl/NRR-2352L Pig 11 of 7 pgyyg l 3.6.1.3 SURVEILLANCE REQUIREMENTS (continuet, SURVEILLANCE FREQUENCY S 3.6.1.3.11 ----.
NOTh---------.2.
nly required to be merin MODES 1 2 .EW edM. .Ilo.c.s. .Me. Mc-hhd ....
l Verify combined leakage rate of 1 gpm In accordance times the total number of PCIVs through with the hydrostatically tested lines that Primary penetrate the primary containment is not Containment exceeded when these isolation valves are Leakage Rate tested at 2 1.1 P,. Testing Program SR 3.6.1.3.12 ------------------NOTE------------------
Only required to'be met in MODES 1,
- 2. and 3.
Verify each cutboard 42 inch primary 18 months containment purge valve is blocked to restrict the valve from opening > 50 .
SR 3.6.1.3.13 - ----------
NOTE-------------------
Not required to be met when the Backup Hydrogen Purge System isolation valves are open for pressure control. ALARA or air quality considerations for personnel entry. or Surveillances or special testing of the Backup Hydrogen Purge System that require the valves to be open.
Verify each 2 inch Backup Hydrogen Purge 31 days System isolation valve is closed.
l 9901140172 DR 990106 4
p ADOCK 05000440%
PDR ))
3.6-19
~
PERRY - UNIT 1 Amendment No. ,-
Att:chment 3 PY-CEl/NRR-2352L Paga 2 of 7 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.1 Offsite Dose Calculation Manual (ODCM) (continued)
- c. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of, or concurrent with, the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made.
Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.
5.5.2 Primary Coolant Sources Outside Containment This program provides controls to minimize 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. The systems include the Low Pressure Core Spray System, High Pressure Core Spray System, I^
' ^j Residual Heat Removal System Reactor Core Isolation Cooling System, hydrogen analyzer por, tion of the Combustible Gas Control h Nhtu. System, Post-Accident Sampling System, and Feedwater Leakage Control Systemj The program shall inc de the following gg, ophi a. Preventive maintenance and periodic visual inspection W' #
A;#+ requirements; and
- sab~ h *f
- b. , s3 s+ 3^
,, Integrated leak test requirements for each system at refueling cycle intervals or less. u To I
FukcAv 5.5.3 Post Accident Samplino This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions. The program shall include the following:
- a. Training of personnel;
- b. Procedures for sampling and analysis; and
- c. Provisions for maintenance of sampling and analysis equipment.
1 (continued)
PERRY - UNIT 1 5.0-7 Amendment No. 69
Attachment 3 PY-CEl/NRR.2352L P g'a 3 of 7 Primary Containment-Operating B 3.6.1.1 BASES (continued)
SURVEILLANCE SR 3.6.1.1.1 REQUIREMENTS Maintaining the primary containment OPERABLE requires compliance with the visual examin6tions and leakage rate test requirements of the Primary Containment Leakage Rate Testing Program. Failure to meet air lock leakage testing (SR 3.6.1.2.1 and SR 3.6.1.2.4), secondary containment bypass leakage (SR 3.6.1.3.9). resilient seal primary cor.tainment purge valve leakage testing (SR 3.6.1.3.6). main steam isolation valve leakage (SR 3.6.1.3.10), or hydrostatically tested valve leakage (SR 3.6.1.3.11) does not necessarily result in a failure of this SR. The impact of the failure to meet these SRs must be evaluated against the Type A. B. and C acceptance criteria of the Primary Containment Leakage Rate Testing Program. The Frequency is required by the Primary Containment Leakage Rate Testing Program.
The Appendix J -9p mptions approved to date cre listed below. Appendix J. Option A exemations that are applicable to Appendix J. Option B may Je utilized for Appendix J. Option B testing, unless they have been specifically revoked by the NRC (Ref. 3). Additionally.
Bechtel Topical Report BN-TOP-1 may be utilized for ILRTs with a duration of less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as noted in Reference 5 and in the Primary Containment Leakage Rate Testing Program,
- a.Section III.D.2(b)(ii) - The air lock seal leakage test of Section III.D.2(b)(iii) of Ap)endix J may be substituted (following normal air locc door opening) for the full-pressure test provided that no maintenance has been performed that would affect the air locks sealing capability (Reference 6).
- b.Section III.D.3 - A one time schedular Exemption was issued to permit Type C testing of certain containment isolation valves to exceed the two year interval. so that these tests could by cr. ducted during the first refueling outage (Reference 7).
(continued) l I
PERRY - UNIT 1 B 3.6-4 Revision No. I l
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AttIchment 3 PY-CEl/NRR 23S?L Primary Containment-Operating l
Pega 4 'wi 7 8 3.6.1.1
\
l BASES SURVEILLANCE SR 3,6.1_._1 (continued) i REQUIREMENTS
- c. Sections Ill.A.1(d). Ill.A.5(b)(2), 111.B.3 and Ill.C.3 - The main steam lines between the inboard and outboard MSIVs (including the volume up to the outboard MS1V before seat drain line valves) are not i recuired to be vented and drained for Type A testing anc
! the main steam line isolation valve leak rates are exempted from inclusion in the overall integrated primary containment leak rate and the combined local leak rate (Reference 8).
- d. Section Ill.D.1(a) - The third Type A test for each 10 year service period is not required to be conducted
! when the plant is shutdown for the 10-year plant Inservice inspection (Reference 8).
e, l
Section III.D.3 - Type C local leak rate testing may l be performed at other convenient intervals in addition l
to shutdown during refueling, but at intervals no greater than 2 years (Reference 8).
T As~La uT@ cu%r 6 wQ Tett leakage prior to 11ie first startup after performing a required leakage test is required to be < 0.6 L for l 1
combined Type B and Type C leakage, and s 0.75 L,,for !
overall Type A leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an l overall Type A leakage limit of < 1.0 L At < 1.0 L, the
- i offsite dose consequences are bounded by,. the assumptions of l l the safety analysis, i
REFERENCES 1. USAR, Section 6.2. i l 2. USAR, Section 15.6.5.
- 3. 10 CFR 50, Appendix J. Option B.
- 4. PY-CEl/NRR 1510L. dated June 24, 1992.
- 5. Letter from NRC (B.J. Youngblood) to CEI (M.R.
Edelman). " Performance of the Preoperational Containment Integrated Leak Rate Test - Perry Nuclear Power Plant. Unit 1." dated June 10, 1985. .
(continued 1 i
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PERRY - I .6-5 l
I Revision No. 1
NCE / RR-2352L Primary Containment-Operating PIge 5 of 7 B 3.6.1.1 BASES REFERENCES (continued) 6. PNPP Safety Evaluation Report Supplement 7. Section 6.2.6 " Containment Leakage Testing." November 1985.
7.
Letter from NRC (T. Colburn) to CEI (A. Kaplan).
" Exemption from 10 CFR Part 50. Appendix J". dated January 22. 1988.
- 8. Letter from NRC (J. Hopkins) to Centerior Services Company (D. Shelton). " Issuance of Exemption from the Requirements of 10 CFR Part 50. Appendix J - Perry Nuclear Power Plant. Unit 1" dated December 4. 1995.
- 9. CT'1O A
f 6 N R,c. b FirW b~,$) b #
oper3W 3 Co.wp aq 3 "EepNA b go c.rt F A 50, Ap(uM y 7 "3 Md- '
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PERRY - UNIT 1 B 3.6-6 Revision No.1 l
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Attzchment 3 PY-CEl/NRR.2352L Pagi6 of 7 PCIVs B 3. 6.1.3 BASES SURVEILLANCE SR 3.6.1.3.11 (continued)
REQUIREMENTS demonstrated at the frequency of the leakage test requirements of the Primary Containment Leakage Rate Testing Program.
This SR is modified by a Note that states these valves are only required to meet the combined leakage rate in MODES 1.
- 2. and 3 since this is when the Reactor Coolant System is pressurized and primary containment is required. In some instances, the valves are required to be capable of automatically closing during MODES other than MODES 1. 2.
and 3. However, specific leakage rate limits are not applicable in these other MODES or conditions.
h^ ?
3R 3.6.1.3.12 Verifying that each outboard 42 inch (1M14-F040 and IM14-F090) primary containment purge supply and exhaust isolation valve is blocked to restrict opening to s 50 is required to ensure that the valves can close under DBA conditions within the time limits assumed in the analyses of References 2 and 3.
The SR is modified by a Note stating that this SR is only required to be met in MODES 1. 2. and 3. If a LOCA inside primary containment occurs in these MODES, the purge valves must close to maintain containment leakage within the values assumed in the accident analysis. At other times when the purge valves are required to be capable of closing (e.g.,
during movement of irradiated fuel assemblies in the primary containment), pressurization concerns are not present, thus the purge valves can be fully o)en. The 18 month Frequency is appropriate because the bloccing devices are typically removed only during a refueling outage.
SR 3.6.1.3.13 This SR ensures that the 2 inch Backup Hydrogen Purge System isolation valves are closed as required, or, if open, open for an allowable reason. These backup hydrogen purge isolation valves are fully qualified to close under accident conditions: therefore, these valves are allowed to be open for limited periods of time. This SR has been modified by a (continued)
PERRY - UNIT 1 B 3.6-32 Revision No. 1
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Attichment 3 PY-CEl/NRR-2352L Pags 7 of 7 l
Bases insert; Page B 3.6-32 ,
A second Note states that the Feedwater lines are excluded from this particular hydrostatic (water) testing program. This is because water leakage from the stem, bonnet and seat of the third, high integrity valves in the feedwater lines (the gate valves) is controlled by the Primary Coolant Sources Outside Containment Program (Technical Specification 5.5.2)). The acceptance criteria for the Primary Coolant Sources Outside l Containment Program is 5 gallons per hour.
i
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k, l
s l
J TABLE 1.8-1 (Continued)
Requlatory Guide (Rev.;RRRC Category) Decree of Conformance Reference .j f
1.149 - (Revision 1 - 4/87) [
l Simulation facilities for use in PNPP conforms to this guide. -
operator license examinations at nuclear power plants. l
-[
1.150 - (Revision 1 - 2/83) t Ultrasonic testing of reactor vessel PNPP conforms to the alternative method -
welds during preservice and inservice presented in Appendix A of this guide, f' examinations g 1.155 - (Revision 0 - 6/88)
Station Blackout PNPP conforms to this guide. 15.8.2, 15H r
$ l 1.163 - (Revision 0 - 9/95) f Performance-Based Containment Leak-Test PNPP complies with this guide with the Program following exceptions:
t BN-TOP-1 methodology may be used for ;
Type A tests. gTg i E ch The corrections to NEI 94-01 that are identified on the Errata Sheet attached yg 0 ;o p. ;
$@ to the NEI letter, " Appendix J Workshop *[* ;
1$
Questions and Answers," dated March 19, M
.E 1996, are considered to be an integral E eO Part of NE O , ,
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TABLE J.2-1 (Continued)
Q T
I 1 ~i QualiN)
Group Principal (5) 7 Safety (2) I ;
Principal Component (I) Construction Seismic (6) go -
Class Location ( )cation Classifi- Code Category Comment y II. Nuclear Boiler System 2 '
5
- 1. Vessels, level i Instrumentation condensing chambers
- 2. Vessels, air 1 D A III-1 I D ;
T accumulators 3 A/D C F 3. Piping, relief III-3 I --
? valve discharge '3 C/D C h C 4. III-3 I (7)
Piping, main steam, )
within outermost 5.
isolation valve Piping, feedvater 1 A/D A III-1 I fg- 1 within outermost o
" isolation valve )
1 A/D A III-1 I '
. Pipe supports, main .
[
f !
steam 1 D A
- 7. Pipe restraints, III-NF I
!i
- main steam 2 D N/A N/A I
- 8. Piping, main steam, . 7h.
between isolation valve and M.0. 2"s '?
stop valve 2 A B III-2 I fQ ,)
- 9. Piping, main steam between M.0. yjg(pj oEg 1 L stop valve and g "
turbine stop valve NSC A/T g D B31.1 N/A (24) r-i
V V tV TABLE 3.2-1 (Continued) ;> 1 ]
, sT Qualik) '
Group Principal (5)
% 3 Safety I2I Classifi- Construction Seismic (6) y Principal Component ( Class Location ( cation Code Category Coment , S 0
- 2. Heat exchangers, 1T secondary side 3 A C III-3& $pf
,%p)
TEMA-C I
- 3. Piping, within !
outermost isolation i valves 1 C A III-l I (8)
- 4. Piping, beyond outermost isolation P' valves 2 A B III-2 I (8) y 5. Pumps 2 A B III-2 I y 6. Pump motors 2 A N/A None I
- 7. Valves, isolation and LPCI line between 1 C A III-l I (8)
F 8. Valves, isolation, other 2 A B III-2 I
- 9. Valves, beyond isolation valves 2 A B III-2 I
- 10. Mechanical modules 2 M,A,C N/A None I TT>
- 11. Electrical modules $n with safety function 2 M A,C N/A' IEEE I w !I! g
,, 12. Cable with safety $31 yQ function ,
Suppression pool 2 M,A,C N/A IEEE I ^ya p g- 13. g
' g- strainer 2 C N/A II, IX I (36) M G"
E"
L
/. 7 "r' .
TABLE 3.2-1 (Continued) '
1
] .
\}* %
QualikE)
Gr up Principal (5)
. . 3 Safety (2)
' Classifi- Construction Seismic (6)
Principal Component Class Location ( cation Code Category Coment $ _d XIX. Reactor Water Cleanup j System f
- 1. Vessels:
demineralizer Filter /
NSC C C III-3 N/A h:r 3 D ( 1
- 2. Heat exchangers t !
carrying reactor water NSC C C III-3,TEMA-C N/A
- 3. Pump suction piping, !
to outermost
[ isolation valve 1 C A III-1 I (8),(16) 4 4. Pump discharge m piping, to RHR and ..
feedwater 2 M,W B III-2 I (8) j
- 5. Pumps NSC C C III-3 N/A ;
- 6. Valves, isolation valves and piping between 1/2 C A/B III-1,III-2 I (8),(16),(37)
- 7. Valves, pump l ;
y discharge to RHR !
and feedwater 2 M,W B ASME III-2 I (8) !
- 8. Filter /demineralizer NSC C C III-3 N/A [h l
- 9. Filter /demineralizer j, m 3 r gg precoat subsystem NSC C D B31.1 N/A g, @ 2 -
U. S.
~" Eh! f S
8 $
p i l
l
_ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ ._- - _ _ ~ . . . . , , _s, . _ __
t i
I C
TABLE 3.2-1 (Continued) Ig ;
Qualig)
Group Principal ( ) I,2) -
Safety (2) \ 3, Principal Component (U Class Location I ) Classifi-cation Construction Code Seismic Category (6)
Comment I>
F
- 10. Diesel generator combustion air intake, exhaust Tp 7 )
system intake and exhaust systems p3_.
(
intake air filter and valves 3 M,0 C III-3 w 11. Exhaust silencer I
(3 w f* l and crankcase vent -
O
" piping NSC M,0 N/A N/A N/A
- 12. Jacket water. cooling 3 M C III-3 I (S
(i TEMA-C 6
XXX. Power Conversion System
,p ,
r ; .
- 1. Main Steam piping -(See Section II of this Table, Items 7 through 12)
- 2. Main Steam Lines from Turbine Control h'I1 wj Valves to Turbine 6
'j Casing NSC T N/A ANSI B31.1 N/A (20)
- 3. Piping and valves, other NSC A,T,M N/A ANSI B31.1 N/A (21),(22)
Pressure Vessels 4.
5.
NSC A,T,M N/A VIII-l N/A [OSk Condensate &
Feedvater Pumps NSC A,T,M N/A N/A hj N/A yja
- 6. All Other Pumps NSC A,T,M N/A N/A N/A g M
r-
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ __ -_ ___ ,ew' m- w v. ._ _, -
l ..
TABLE 3.2-1 (Continued) Y i Quali[X)
Group Principal (5)
Safety ( ) Construction Seismic (6) . j Principal Component (I) Class Location ( )cation Classifi- Code Category Comment .
'2
- 35. Containment spray f '
Piping and Nozzles 2 C B III-2 I
- 36. Dryvell Vacuum I -
bh 2 C B III, NEMA, Relief #l IEEE l
h1 XXXVI. Other Components j
- 1. Containment Crane 3 C N/A N/A I [
w 2. Refueling Cask Crane 3 M N/A N/A I pi 1
L . Containment #
& isolation valves [
N and piping 2 C B- III-2 I t :.
between for all f o
I containment 2 /
penetrations not listed above gl D
XXXVII. Suppression Pool Q
Make-up System )
I
/ $m9 o, y j
- 1. Valves 2 C B III-2 I
- 2. Piping 2 C B III-2 I yy
- 3. Electrical modules g
- r x V g vith safety function 2 C,M N/A IEEE I
. .E- E 8-w .
_.-_.-._____----__-.._-_-.-----.----__-._.___________-_-_-__-__--____-_.__]
I 9
TABLE 3.2-1 (Continued)
Group I) Principal (5)
Safety (2)
Principal ComponentID Class Location I ) Classifi-cation Construction Code Seismic (6)
Category Comment XLIX. Feedvater Leakage Control System c~ - -
w b,ivislogy
- 1. Piping and valves of the id:::: system 2 A B III-2 I
- 2. Piping and valves of theeud.;arpsystem 2 A B III-2 I ta y Di d s l ow 1
$ ^
2'k S$~e$
R E e, A
13 n
r-
f f '" - -
% ;s. p pro &A Gr ink TABLE 3.2-1 (Continued) Attachment 4 PY-CE1/NRR-2352L Page 8 of 74 UL Underwriters' Laboratories, Inc.
! UL 507 l " Safety Standards for Electric Fans" (ANSI)
UL 586 "High Efficiency, Particulate, Air Filter Units" UL 900 " Air Filter Units" IEEE The Institute of Electrical and Electronics Engineers, Inc.
10 CFR 50 Title 10, Code of Federal Regulations, Part 50,
" Licensing of Production and Utilization Facilities"
- 6. I - Constructed in accordance with the requirements of Seismic Category I structures and equipment as described in Section 3.7, Seismic Design. N/A - The seismic requirementy are not applicable to the equipment.
- 7. Safety relief valve discharge line piping from the safety relief
' valve to the suppression pool consists of two parts. The first is attached at one end to the safety relief valve and attached at its other end to the structural steel just below the main steam header through a pipe anchor. The main steam piping, including this portion of the safety relief valve discharge piping, is analyzed as a complete system. The second part of the safety relief valve discharge piping extends from the anchor (located below the mainsteam header) to the suppression pool. Because of the upstream anchor on this part of the line, it is physically decoupled from the main steam header and is, therefore, analyzed as a separate piping system.
- 8. a. Lines 3/4 inch and smaller which are part of the reactor coolant pressure boundary are Safety Class 2 up to and including the root or isolation valve.
- b. Instrument lines larger than 3/4 inch which are connected to Safety Class 1 (SC-1) process lines have a restricting orifice installed between the process connection and the root or isolation valve.
- c. Lines that are connected to safety class process lines are
- classified as the same safety class as the process line from the process line connection to and including the root or isolation valve except as noted in paragraph a above.
3.2-63
preef Aed Or Iaf5%hM i noO6^p5, Y CE RR-2352L Page 9 of 74 TABLE 3.2-1 (Continued) 21.
A certification is obtained from the vendors gf the turbine stop valves and turbine bypass valves which shall certify compliance with the following:
a.
All cast pressure-retaining parts of a size and configuration for which volumetric examination methods are effective are examined by radiographic methods by qualified personnel.
Ultrasonic examination to equivalent standards may be used as an alternate to radiographic methods.
Examination procedures and acceptance standards are at least equivalent to those specified as supplementary types of examination in ANSI B31.1, 1967, Paragraph 126.4.3.
b.
The vendor of the turbine stop valves and turbine bypass valves utilizes quality control procedures equivalent to those defined for valves in the GEZ 4982A, 22.
In addition to a piston-type check valve inside the dryvell and a
/ piston-type check valve outside containment, a third valve with high leak-tight integrity is provided in each feedvater line outside containment. Buffer pistons vill be provided in the feedvater line check valves. These buffer pistons vill slov valve closing to limit check valve slam and associated pressure transients. The high leak-tight integrity isolation valve vill be remote-manually operated from the control room using signals which indicate loss of feedvater flow.
The classification of the feedvater lines from the reactor vessel to and including the check valve outside containment is Safety Class 1; from this check valve to the high integrity isolation valve (see class it is Safety Class 2; beyond the third valve it is nonsafety Figure 10.1-3).
23.
A nonsafety structural steel annex forms part of the radvaste building. See Section 3.8.4 for further discussion.
! 24.
Joseph M. Hendrie (Deputy Director for Technical Review, Director of Licensing, USAEC) to John A. Hinds, (Manager, Safety and Licensing, General Electric Company) letter of April 19, 1974, provides an appropriate standardized approach to MSL and MFL classification, which is acceptable as an alternate to the guidelines currently specified in Regulatory Guide 1.29 (Revision 3/9/78).
3.2-70
Attachm:nt 4 PY-CElrNRR-2352L Page 10 of 74 C-5 l -G4l M I
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S PERRY NUCLEAR POWER PLANT THE CLEVELAND ELECTRC ILLUMINATWG COMPANY Residual Heat Removal System Figure 5.4-13 (Sheet 2 of 3) I (Dwg. D-302-642)
a Attachment 4 PY-CEl/NRR-2352L Page 11 of 74 M\S CE l FEEDWATER'
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@ PERRY NUCLEAR POWER PLANT THE CLEVELAND ELECTRIC ILLUMINATNG COMPANY Residua 1 Heat Removal System Figure 5.4-13 (Sheet 2 of 3)
(Dwg. D-302-642)
Attachment 4 PY-CEl/NRR-2352L Page 12 of 74 In secondary the order tocontainment minimize the amount of radioactive that leaks to material following a design basis accident, containment penetrations are provided with redundant all primary
, ASME Code,Section III, Class 2, Seismic Category I isolation v l the primary containment a ves, one inside of
- tier af pipe betverr and one outside of the shield building I A. %^ 3 ASME Code,Section III, Class the t.m 2. antsin.. cat i=h t i n r.1 :: is also functions to p:r. renLir m.6 This isolation valve arran A
a ngle r 2 1.. e, .m "through-line" leakagep. :=:A in4 l't eat w ee +eJL % gemen tWJ-t A fu of 2ny_
m ~_gh ii # 17'uge 6 m'
- 5 f.ECu i e g n-c The containment ta .k.hi t 6 2 45 ro Section k. k m%ul containment isolation system is discusse(pn . . .
4 kgM' The Section 7.3.1. and reactor vessel isolation control system is discus sed in The containment boundary and all penetrations except with guard pipes terminate in the annulus. for penetrations leakage and penetration leakage are considered to be tTherefore, c the annulus. otally directed to The sources listed in Table 6.2-33 are a summary of potential leakage paths that could bypass the AEGTS .
design basis accident leakage is 0.2 percent by w i h The containment atmosphere in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. e g t of the contained the sources listed in Table 6.2-33 is 5 04 percThe maximum .
containment leakage. ent of the total commitment This value vill be the technical specification Specifications.for leakage bypassing the AEGTS as listedn in cal the Tech i In order to verify that bypass leakage vill be within this limit, the total amount of potential ;
a testing and evaluation program vill be conducted on isolation valves, personn e l
guard pipes as described in Section 6.2.4.3.1 airlocks and The expected leakage rates per valve have been c l on Table 6.2-33 for the potential bypass leakage calculations, s. patha culated and are sho it was assumed that In these vill be the same as the shop test the onsite leakage limit per valve specifications. limits given in the valve 6.2-70 Revision 1 March, 1989
Att chm nt 4 PY-CEl/NRR-2352L Page 13 of 74 insert Page 6.2 70
. .or some other acceptable configuration such as a closed system outside of containment. The piping out to the outboard containment isolation valve or in the closed system.
l l
l I
1 l
l l
[
Attachm nt 4 PY-CEl/NRR-2352L Paga 14 of 74 rapid valve closure under operating conditions, is required by the design specifications for piping systems associated with containment isolation valves. The valve operability assurance program for active, safety-related valves is discussed in Section 3.9.
Electrical redundancy is provided for power operated valves. Power for the operation of two isolation valves in a line (inside and outside containment) is supplied from two redundant, independent power sources without cross ties. In general, isolation valves outside containment are powered from the Division 1 power supply while isolation valves within containment are povered from the Division 2 power supply. Both Division 1 and Division 2 valves are generally povered from ac power sources. Loss of power to each motor operated valve is annunciated.
Provisions for detecting leakage from remote-manually controlled systems are discussed in Section 5.2.5. Detection of leakage from containment is discussed in Section 6.2.6. Section 6.7 describes the main steam isolation valve leakage control system.
The fraction of the total containment leakage following a design basis accident /LOCA that could bypass the containment annulus exhaust gas treatment system is limited to the leakage from sources which constitute open systems or nonsafety-related systems. Safety class systems which are open systems are considered. For example, it is assumed, for onsafety that the only portio of such a system remaining ' 11 "Mg : ::!:rie ould be the containment isolation piping between these valve . Additional leakage sources considered are the containment penet tions with guard pipes. #
fo e Ae-te.rM ab' c n of pdmea \ 9c.
m \s- % pr % p Co w6r h n h FuAwah d h , N e
.por p hw\) o ne.- hoar A lt w i $ s L.oc, r) ',
6.2-82 l
. .- .. - -. .. . . - _ . _ . - - -- - ~~ -. - - - - . . ~ . . . -
AttachmInt 4 PY-CEl/NRR-2352L Page 15 of 74 The containment boundary is surrounded by the annulus. Containment penetrations, except those with guard pipes, terminate in the annulus.
Therefore, containment shell leakage and penetration leakage are totally directed into the annulus.
Containment design basis accident leakage is 0.2 percent, by weight, of the contained atmosphere in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The maximum permitted leakage rate from potential sources listed in Table 6.2-33 is 6.72 percent of the total containment leakage. The maximum allowable combined test leakage rate from potential sources listed in Table 6.2-33 is 5.04 percent (0.75 times 6.72 percent) of the total containment leakage.
This value is the technical specification commitment for leakage bypassing the containment annulus exhaust gas treatment system.
To verify that the total amount of potential bypass leakage is within the established limit, the following test and evaluation program will be conducted in accordance with the Containment Leakage Rate Testing Programs
- a. Isolation valves F' d ad IhT he WW .A ASb'o -
W e-3 5"
,G Since it is assume that nonsafety-related systems outside the containment isolation valves will not remain intac Jc;1:1ir.; e j containment atmosphere must terminate at the outer contai isolation valve seat. The same effect is possible for open - i l --fety class systems. To assure that this s as potential source o ea, age is checked, isolation valves listed in Table 6.2-40 are included in the periodic " Type C" test program discussed in Section 6.2.6. The test method is also described within this section. By measuring the time related pressure decay or by directly measuring the leakage flow rate, each valve is quantitatively evaluated for leak tightness
- --- rr en O (gg %n h F ack w hk k i uJ j bF bM '"" #"D h#
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_ Revision 9 6.2-83 April, 1998 I
, ^r ,r f
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Attachment 4 Such alternates are described in Sections 6.2.4.2.2.1 through 6.2.4.2.2.3. The final measure by which GE is assured that the BVR f[o$2M2L design is in agreement with the CDCs is receipt of the Advisory Committee on Reactor Safeguards (ACRS) letters permitting construction and operation of previous plants with comparable valving arrangements.
6.2.4.2.2.1 Justification with Respect to General Design Criterion 55 The reactor coolant pressure boundary, as defined in 10 CFR 50, Section 50.2 (v), consists of the following: reactor pressure vessel; pressure retaining appurtenances attached to the vessel; valves, and pipes which extend from the reactor pressure vessel to, and including, the outermost isolation valve. The lines of the reactor coolant pressure boundary which penetrate containment are capable of isolating the containment, thereby precluding any significant release of radioactivity. Similarly, for lines which do not penetrate containment, but which do comprise a portion of the reactor coolant pressure boundary, the design ensures that isolation of the reactor coolant pressure boundary can be achieved. Items a, b and c, belov, address influent lines, effluent lines and conclusions, respectively, with regard to GDC 55.
- a. Influent Lines Influent lines which penetrate containment and the dryvell directly to the reactor coolant pressure boundary are equipped with at least two isolation valves. One valve is inside the dryvell; the second is as close as possible to the external side of containment. These isolation valves protect the environment. Where needed, protection of the containment in the event of pipe rupture outside the dryvell but within containment is further ensured by extension of the dryvell by use of guard pipes. These guard pipes, together with the isolation valves, assure protection in the event of an active 6.2-87 l
l
l AttIchment 4 PY-CEl/NRR-2352L P g)17 of 74 ;
failure between dryvell'and containment. Table 6.2-34 lists those influent lines that comprise part of the reactor coolant pressure boundary and penetrate containment. The purpose of this table is to summarize the design of each line with respect to the requirements of GDC 55. Items 1 through 8, below, demonstrate that, although a word-for-word comparison with GDC 55 is not always practical, it is possible to demonstrate adequate isolation provisions on some other defined basis. %
(-
( QuMW*W
- l. Feedvater P121/I'll2 and P414/P410)
Feedvater lines are part of the reactor coolant pressure ;
boundary since they penetrate both the containment and dryvell and connect to the reactor pressure vessel. Each line includes three isolation valves and is enclosed in a guard pipe.
The isolation valve inside the dryvell is a control closure anti-vater hammer check valve. The first isolation valve outside containment is also a control closure check valve and is located as close as possible to the outside containment vall. The outermost valve is a motor operated gate valve.
The two control closure check valves are designed and tested
! to close with no reverse flo
]:n '
Extension of the dryvell by means of the guard pipe protects the containment from overpressurization in the event of a feedvater line break between the dryvell and containment valls. The internal design temperature and pressure for the guard pipes which enclose the feedvater lines are the same as the design values specified for the enclosed feedvater lines.
6.2-88
m Atttchment4 LSAIk ,PY-CEl/NRR-2352L Page 18 of 74 Should a break occur in a feedvater line, th control closure check valves prevent significant loss of re tor coolant inventory and provide immediate isolation. The outermost motor operated valve does not close automatically upon occurrence of a protection system signal since, during a LOCA accident m=4ntenance of reactor coolant makeup from all
,1 rL@ne a sa-o n+ y 6. = r c. u p sources , s aesiracle. Inis valve, riovever, can be remotely hi closed from the control room to provide long term, h e protection when, in the judgment of the operator, continued makeup from the feedvater system is no longer necessary. In 1^ M addition, after feedvater flov terminates, the operator vill initiate the feedvater leakage control system (refer to Section 6.9) to provide a positive water seal on the 6 1-+i-~
- E ::.
u.rt
- 2. High Pressure Core Spray Line (P410/P411)
The high pressure core spray line penetrates both the containment and the dryvell and connects to the reactor pressure vessel.
Isolation is provided by a hydraulically testable check valve inside the dryvell and a motor operated gate valve as close as possible to the outside of the containment vall. This gate valve maintains long term leakage control. Position indication for the hydraulically testable check valve is provided in the control room. The gate valve is automatically and remote-manually operated. A guard pipe is not necessary since the high pressure core spray fluid is at an energy level during system operation that containment overpressurization cannot result should the line break between the containment and the dryvell.
6.2-89
Attachment 4 PY-CEl/h!RR-3352L
- Page 19 of 74 I
i l'
insert Page 6.2-89
, i l These check valves are tested in accordance with Technical Specification 5.5.6, j inservice Testing Program, to verify this closure function. An Appendix J exemption for these containment isolation check valves documents that they utilize an alternate testing methodology involving visual inspection of the valve seats per the inservice Testing Program to verify their proper closure.
insert Power to these motor operated valves can be provided from an attemate division under .
administrative controls, if necessary following a LOCA.
Insert
... seat, stem and bonnet of the motor operated valve in each line. The check valves, coupled with the single motor-operated high integrity leakage protection gate valve on each line, provides an acceptable configuration for the feedwater lines.
A branch line connects to the Feedwater line outboard of the second Feedwater check valve, which is outboard of the containment. This branch line provides the pathway for l RWCU water and RHR shutdown cooling water to retum to the reactor vessel. For the RHR shutdown cooling retum line, a safety-related globe valve is treated as a high integrity containment isolation valve, similar to the Feedwater gate valves. The RHR system " outboard" of the globe valve is also treated as a closed system outside of containment, to control any leakage. For the RWCU retum line, the piping " outboard" of the RWCU branch line check valve leads directly back to containment penetration P132, and is ASME Code Class 2, Seismic Category 1, protected from pipe whip, missiles and jet forces, and analyzed for " break exclusion". This line is also treated as a closed <
system outside of containment (see Table 6.2-40 for testing details on containment penetrations). The design of these branch lines also provides an acceptable configuration.
(
l l
l l
_.-.__.,.u,-- n,. y ,
Attachment 4 PY-CEL/NRR-2352L Pigs 20 of 74 vessel. Isolation is provided by a check valve inside the dryvell and a check valve and explosive valve outside the l dryvell. The explosive valve provides an absolute seal for long term leakage control, as well as preventing leakage of sodium pentaborate into the reactor pressure vessel during normal reactor operation. Since the standby liquid control line is normally an isolated, nonflowing line, rupture is extremely improbable. However, should a break occur subsequent to actuation of the explosive valve, the check valves ensure isolation.
,;- 7.
Residual Heat Removal Shutdown Cooling Return Lines (P121/P112 and P414/P410)
The residual heat removal shutdown cooling return lines discharge into the feedvater line between the testable check t
valve and the motor operated gate valve outside of containment. A check valve and a normally closed, motor operated, remote-manually actuated globe valve provide for isolation of the residual eat remov sh own c ling return lines. S*- Y'^ l' "b*** For auth oai disuss
- 8. Reactor Vater Cleanup System Line (P419/P432)
The discharge line from the reactor water cleanup pumps penetrates containment and serves the reactor water cleanup regenerative heat exchangers inside containment.
Automatically actuated motor operated gate valves, one inside, one outside containment, provide for isolation.
I l,
4 i
6.2-93
l
~ l
%, gg p u& a-w - % A. 3 1
Attachment 4
- b. Effluent Lines PY-CEl/NRR 2352L Page 21 of 74 Effluent lines that form part of the reactor coolant pressure I boundary and penetrate containment and/or the drywall are equipped with at least two isolation valves, one valve is inside the drywell, the other outside, but as close as possible to the containment. Where needed, the containment is protected, in the event of a pipe rupture outside of the drywell but inside containment, by guard pipes which enclose the process lines, forming an extension of the drywell. This combination of isolation valves and guard pipes assares protection in the event of a failure between drywell and containment walls.
Table 6.2-35 lists those effluent lines that comprise part of the reactor coolant pressure boundary and that penetrate containment and/or the drywell. Items 1 through 4, below, address specifies of 1 these lines.
- 1. Steam Lines (P124/P116, P416/P414, P122/P115, and P415/P415)
Steam lines include main steam, main steam drain, residual i 1
heat removal, and reactor core isolation cooling steam lines.
l The main steam lines from the reactor pressure vessel to the turbine penetrate both drywell and containment. Main steam l
line drains (one for each main steam line) in the drywell are '
headered together to form one line which penetrates both drywell and containment. Isolation for the main steam lines and main steam drain line is provided by automatically actuated block valves, one inside the drywell and one outside containment except for that provided by the outboard MSIV drain valves, 1B21F067A, B, C, D, which are locked closed.
l l
l Revision 9 6.2-94 April, 1998 l
_ ,_ - ,__ - - - - - - - - - - ^ ~ ~ - " ~ ~
9, sl & b r- En fo rA ch " N* S fS tachment 4 PY-CEl/NRR 2352L i The residual heat removal steam supply and reactor core Page 22 of 74 isolation cooling turbine steam line branches from the main steam line inside the dryvell. Isolation for this line is provided by one normally open gate valve and one normally closed globe valve inside the dryvell, and one normally open gate valve outside containment. These motor operated valves are capable of automa tic and remote-manual actuation.
Use of guard pipes to enclose these steam lines prevents containment overpressurization in the event of line break between the dryvell and containment valls. The internal l
design temperature and pressure for the guard pipes which '
enclose these steam lines are the same as the design values specified for the. enclosed lines.
- p. 2. Reactor Water Cleanup Lines (P131/P132)
The reactor water cleanup pumps are located outside containment; the heat exchangers and filter demineralizers, are located inside containment, but outside the dryvell. The reactor water cleanup pump suction line from the reactor recirculation system lines and the reactor bottom head penetrates the dryvell and containment. Two automatically actuated, motor operated valves provide for isolation of this l
line. One valve is just inside the dryvell, the other is l
l outside containment. A guard pipe encloses the line between the dryvell and containment valls.
L l The reactor water cleanup pump discharge line to the heat exchangers and filter demineralizers penetrates containment.
Two automatically actuated, motor operated valves (one inside i and one outside containment) provide for isolation of this line.
I 4
6.2-95
m ._ ._ _ . ._. _ _ . _ _ . _ _ _ _ _ . - _ . _ . _ . _ _ _.._ .._ _ _ _ . _ _ _
~
Attachmsnt 4 6e isfe rM' % - su h
$ ; ,, p v ; A.4A
- yS 2o A blowdown line from the filter demineralizers penetrates containment and divides to form separate lines to the condenser and radvaste system. Automatically actuated, motor operated valves, one inside and one outside containment, !
provide for isolation of this line.
The return line from the filter demineralizers penetrates t-
- E containment and connects to the feedvater line between the u outboard feedvater gate veP a eid the outboard (feedvater) check valve. Two automaticaay actuated, motor operated gate valves provide for isolation of this line. One valve is '
inside, the other outside of containment.
i l
- 3. Residual Heat Removal Shutdown Cooling Line (P421/P406) l The residual heat removal shutdown cooling line branches from the B reactor recirculation loop and penetrates both the t dryvell and containment. A check valve, in parallel with a normally closed, remote-manually actuated, motor operated valve isolates the line inside the dryvell; a normally closed, remote-manually actuated, motor operated valve isolates the line outside containment. A guard pipe encloses this line l
l from the dryvell vall to the containment vall to protect against containment overpressurization in the event of a line break.
l
- 4. Recirculation System Sample Line (NA) l A sample line from the recirculation system penetrates the dryvell. This line is 3/4 inches in diameter and is designed in accordance with the requirements of the ASME Code,Section III, Class 2. A sample probe with a 1/8 inch diameter hole is located inside one recirculation discharge line within the dryvell. In the event of a line break, this probe acts as
- 6.2-96 I l - - - .
C "
7"e Attachrn nt4
%5 i ro vIELL A 'fc r~ Itefe % hcA 1 oc LM S PY-CEl/NRR-2352L
{A- Page 24 of 74 a restricting orifice and limits escaping fluid flow. Two air operated valves which fail closed are provided for isolation of this line. Both sample isolation valves are located outside the dryvell.
- c. Conclusions Concerning General Design Criterion 55 To assure protection against the consequences of accidents involving the release of radioactive material, piping which forms portions of the reactor coolant pressure boundary has been shown to provide adequate isolation capability on a case-by-case basis. In all cases, a minimum of two barriers is shown to protect against release of radioactive materials. Where necessary to protect the containment against overpressure, guard pipes are provided which enclose the process pipes between the dryvell and containment valls.
In addition to satisfying the requirements of GDC 55, the pressure retaining components which comprise the reactor coolant pressure boundary are designed te satisfy other appropriate requirements which minimize the pro aility or consequences of an accident rupture. Quality requirev 1ts for these components ensure that they are designed, fabrice.,ed and tested to the highest reactor plant component standards. The classification of components which comprise the reactor coolant pressure boundary is Quality Group A; these components are des igned in accordance with the ASME Code,Section III, Class 1.
Additional infon.ation concerning classification is presented by Table 3.2-1. The containment and reactor vessel isolation control system is addressed in Section 7.3.
6.2-97
PY-CEl/NRR-2352L Pigs 25 of 74 {
l Q :s ('a fcovsN W I^SMN^ ~ N' '
i i
d.
Specific activity in the reactor coolant was conservatively assumed to be 6.56 pCi/g of I-131 and 34.9 uci/g of Xe-133, with other ;
isotopes in proportionate quantities. This corresponds to spike conditions. !
e.
Turbulence resulting from the high blovdown rates and operation of fan coolers in containment was assumed to ensure good mixing in the entire containment volume.
f.
Containment air was assumed to be released through two 18 inch purge lines, one supply and one exhaust, for five seconds.
Constant flow rates through the open purge lines corresponding to the maximura containment pressure of approximately 3.0 psig during the release period (see Figure 6.2-2) were used to determine a total flow to the environment of 1,020 pounds. This value is conservative since it ignores lover flow rates due to lover containment pressures and partial closure of the purge isolation 1 valves at times prior to five seconds. i'
- g. No credit was allowed for iodine removal by the 99 percent efficient charcoal adsorbers in the containment purge exhaust lines.
I
- h. Site boundary X/0 (see Table 15.6-12) was used in the dose calculation.
6.2.4.3 Design Evaluation t
6.2.4.3.1 General Evaluation To ensure the accomplishment of the design objective stated in Section 6.2.4.1, redundancy is provided in all design aspects of the containment isolation systems. Mechanical components are redundant and each isolation valve is protected, by separation and/or adequate 6.2-106 l
I l
l __ ._ ___ _ _ _ _ _ _ _ _ _ _ _ _ _ __
w mm 9
i CONTAINMEN1 g3) CDC/ Line Fig. Sys. and Penetration No2 Reg. Size UNIT 1 UNIT 2
- Essen{jgt 6.2-60 valve t Cuide System Number Fluid (in) Sys. Arr. No. Number [g P120 P427 GCC56 Containment Leak Rate - Air 8 No 35(b) Spect. Flange o Pressurization Line Air 6 No 35(b) Spect. Flange I Air 8 No 35(b) Blind Flange 1 P121 P112 GDC55 Feedwater A, RHR, and Water 20 Yes* 2 B21FD65A 0 RWCU Return to Reactor Water 20 Yes* 2 B21F032A ' O Pressure Vessel untar 20 Yes* 2 N27F559A t P122 P115 GDC55 Main Steam Line C (Whr arcam iz.
to Ne4 Yes*
L 1(a) e EJ 2-F o F3 A__ O_ -
821F028C 0 Steam 26 Yes* 1(a) 821F022C I Steam 1 1/2 Yes* 1(a) 821F067C 0 Steam 2-1/2 Yes* 1(a) E32F001J 0 P123 P117 GDC55 RCIC Pump Discharge and Water 6 Yes* 5 E51F066 t RHR Head Spray Water 6 Yes* 5 E51F013 0 Water 6 Yes 5 E12F023 0 P124 P116 GDC55 Main Steam Line A Steam 26 Yes* 1(a) B21F028A 0 Steam 26 Yes* 1(a) 821F022A 1 Steam 1-1/2 Yes* 1(a) 821FD67A 0 Steam 2-1/2 Yes* 1(a) E32F001A 0 P131 P132 GDC55 RWCU Pump Suction Water 6 Yes* 49 G33F001 I Water 6 Yes* 49 G33F004 0 P132 P408 GDC55 RWCW Line f rom Regenerative Water 6 Yes* 44 G33F040 1 Heat Exchanger to Feedwater Water 6 Yes* 44 G33F039 0 P201 P218 GDC56 Drywell Atmosphere Drywell Atitus. 1 Nc 52 D17F079A 0 Radiation Monitor Line Drywell Atmos. 1 No 52 D17F079B I Drywell Atmos. 1 No 52 D17F071A 0 Drywell Atnos. 1 No 52 017F071B I P203 P301 GDC56 Fuel Pool Cooling Supply Water 8 No 26(a) G41F100 0 Water 8 No 26(a) G41F522 I P204 P302 CDC55 Control Rod Drive to Reactor Condensate 2-1/2 Yes* 3 C11F083 0 Pressure Vessel Condensate 2-1/2 Yes* 3 C11F122 1 P205 P205 GDC56 fuel Transfer Tube Water 24 No 36 Double gasket 1 P208 P122 GDC56 Containment vacuun Atmos. 24 Yes 19 M17F025 0 Relief Atmos. 24 Yes 19 M17F020 I P210 P206 GDc56 Carbon Dioxide to Fire CO 4 2
W 42 P54F340 0 Protection System CO 4 No 42 P54F1098 I 2
P301 P222 GDC56 Fuel Pool Cooling Water 10 No 26(b) G41F145 0 Return Water 10 No 26(b) G41F140 I P302 P211 GDC56 Backup Hydrogen Drywell Atmos. 2 Yes 39(b) M51F110 0 Purge System Drywell Atmos. 2 Yes 39(b) M51F090 I j e-- a.
e l
Attachment 4 PY-CEl/NRR-2352L Page 26 Of 74 2-32 (Continued)
ATION VALVE
SUMMARY
(1)(2) t44 (2h g
ype Actuation Vatve Position Pwr. horm.
C Pipe Valve M de esis Length (6) Type Oper. Pri. h Norm down Acc.
Shut Post Pwr Fall g7) Isolatg Closure Sourc ' #"
Signal Time (sec) 9) 1EBus[10) Dir.
o 12'-9-1/2" - - - - - - - - - - -
In 16) o NA - - - - - - - - - - -
In 16) o NA - - - - - - -
In
OP CL OP or CL Reh. Flow pe. NA In Chk P P -
OP CL OP or CL -
G 4 e lpe. M Pev. Finw - - -
In (s. M ' 7 +3 " E. m (_t OP ( t. AL Mk . A , [ _33 1 h 3 es~ 178-2-3/4" Globe A A SP OP CL u FC C,E,f,S,N,P,RM (13) Out ce NA Globe A A SP OP CL CL FC C,E,F,S,N,P,RM C
(13) -
Out es 25'-2-3/4" Globe M M -
C - -
Out es 21'-2-3/4" Gate EM l E M CL CL OP Al RM *,HH,11,JJ,KK,LL 22 1 Out C
es pf NA Chk P (11) - CL CL OP or CL -
Rev. Flow ; s i-[ dm In es 43'-6" Gate EM E M CL CL OP or CL Al RM ,EE,FF '"3$g[d],,{
9 1 In es 50'-5" Globe EM E M Cl Cl OP or CL Al A , I4, U , RM, Std i !r es 16'-5 7/8" Globe A A SP OP CL CL FC C,E,F,S,N,P,RM (13) -
Out es Globe A #
, 3L M ggjM O ,
NA A SP OP CL CL FC C,EF.S,N,P,RMf Out es 23 ' 7/8" Globe M M -
- p'e ~
Out l es 20'-5 7/8" Gate EM E M CL CL OP Al RM *,HH,11,JJ,KK,LL 22# TOP * '1 Out es NA Gate EM E M OP OP CL Al L,B,F,H,Y,RM 20/15 2 Out es 148-0" Gate EM E M OP OP CL Al L,B,F,H,W,Y,kM ,RM 20/15 1 Out f
es NA Gate EN E M OP OP CL Al L,B,F,H,RM 27 2 Out es 10'-9" Gate EM E M OP OP CL Al B,F,H,L,RM c 27 1 Out es <10' Globe S E -
OP OP CL Al B,G,RM <3 1 In es <10' Globe S E -
OP OP CL Al B,G,RM C
<3 2 In es <10' Ball EM E M OP OP CL FC B,G,RM C
<3 1 Out es < '. 0 ' Ball EN E M OP OP CL FC 8,G,RM C
<3 2 Out es 108-9" B' fly EM E M OP OP CL Al B,C,RM 35 1 In es NA Chk P P -
OP OP CL Rev. Ffow In les 188 0" Cate EM E M OP OP CL Al RM Std. 1 In es NA Chk P P -
OP OP CL -
Reh. Flow - -
!n o NA - - = = - - - - - - - -
16) es 2'-6" B' fly EM E M OP OP CL Al B , G, L RM 5 1 In es NA Chk P V -
CL CL CL -
Rev. Flow - -
In es 128-6" Gate EM E M CL CL CL Al B,G,RM es 20 1 In NA Chk P P -
CL CL CL -
Rev.Ffow - -
In es 138 0" B' fly EM E M OP OP CL Al B,G,RM 35 1 Out es NA B' fly EM E M OP OP CL Al 8,G,RM C
35 2 Out es 18' 0" Globe EM E M CL OP CL Al C,G,GG,RM Std. 1 Out or CL C es NA Clobe EM E M CL OP CL Al C,G,GG,RM Std. 2 Out Revision 9 6.2-203 April, 1998
~
t
.-u 4
TA CONTAINMEN GDC/ Line Fig.
Penetration No g3) Reg.
Sys. and Size Essen((gt 6.2-60 Valve UNIT 1 UNIT 2 Guide System Nunber Fluid (in) Sys. Arr. No. Number (g P413 P124 CDC56 PASS Water 3/4 No 61 P87F049 1 Water 3/4 No 61 P87F055 o Water 3/4 No 61 P87F046 I Vater 3/4 No 61 P87F052 0 P414 P410 GDC55 Feedwater B, RHR and Water 20 Yes* 2 B21F065B o RWCU Return to Reactor Water 20 Yes* 2 B21F0328 o Pressure Vessel untar _20 Yes* 2 N27F5598 1 P415 P415 GOC 55 Main Steam Line D
( U3stst' dream et 26 M
Yes*
2- E st Foss s o 1(a) B21F0260 0 Steam 26 Yes* 1(a) 821F022D 1 Steam 1-1/2 Yes* 1(a) 821F067D 0 Steam 2-1/2 Yes* 1(a) E32F001N O P416 P414 GDC55 Main Steam Line B Steam 26 Yes* 1(a) 821F0288 0 Steam 26 Yes* 1(a) B21F022B I Steam 1-1/2 Yes* 1(a) B21F067B o Steam 2-1/2 Yes* 1(a) E32F001E O P417 P128 CDC56 Drywell and Containment Water 3 No 40 G61F080 0 Equipment Drain Sump to Water 3 No 40 G61F075 I Radwaste Water 3/4 No 40 C61F0655 I
. P418 P127 GDC56 Drywell and Containment Water 3 No 43 Floor Drain Sump to G61F170 0 Water 3 No 43 G61F165 I Radwaste Water P419 P432 GOC 55 RWCU Pump Discharge Water 4 No 48 G33F054 0 Water 4 No 48 G33F053 I P420 P412 GDC55 RWCU Backwash Transfer Water 4 No 46 G50F277 0 Pump to Redwaste Water 4 No 46 G50F272 I P421 P406 GDC55 RHR Reactor Shutdown Water 20 Yes 20 E12F008 0 Cooling Suction Water 20 Yes 20 E12F009 I Water 3/4 Yes 20 E12F550 I P422 P407 GDC55 RHR and RCIC Steam Steam 10 Yes* 1(c) E51F063 1 Supply Steam 10 Yes* 1(c) E51F064 0 Steam 1 ies* 1(c) E51F076 1 P4?3 P129 CDCSS Main Steam Line Drain Water 3 No 1(b) 821F019 0 Water 3 No 1(b) 821F016 I P424 P420 GDC55 RWCU to Main Condenser Water 4 ho 14 G33F034 0 and Radwaste Water 4 No 14 G33F028 I Water 1/4 No 14 G33F0646 I sv
Attachment 4
, PY-CEl/NRR-2352L 6.2-32 (continued) Page 27 Of 74 DLATION VALVE SUWARY(1)(2) go No (20 Type Actuation Valve Position Pwr. Norm.
C Pipe Valve Mode ""
Tig length (6) Type Oper. E h Norm down Shut Post Pwr fall (7j isolatg Closure Sourc Time (sec)g) 1EBus(10)
Acc. Sional Dir.
Y1s NA Globe S E -
CL CL OP or CL FC RM <3 -
Out Yes <10' Globe S E -
CL CL OP or CL FC RM P <3 -
Out Yes NA Globe S E -
CL CL OP or CL FC RM P <3 -
Out Yes <10' Globe S E -
CL CL OP or CL FC RM P <3 -
Out 2,O Yrs W i *T' Y Cate EM E M OP CL OP or CL Al RM Std. 1 Q C)
In NA Chk P P OP CL OP or CL Reh. Flow In A NA Chk P P -
OP CL OP or CL - __ Rev. Fim - -
In
._Me4 J9'- 7 8/s a Globe h E A c.L op c4. M AmJ A u -- 33 2 Te )
Yes 16'-P T/6" Globe A A SP OF CL CL FC C,E,f,S,W,P,RM (13) -
Out Yss NA Globe A A SP OP CL CL FC C,E,F,S,N,P,RM C
(13) -
Out Yce 23'-5 7/8" r, lobe M M -
- Out l YIs 20'-5 7/8" Gate EM E M CL CL OP Al RM *,HH,II,JJ,KK,LL 22 1 Out VIs 17'-4-3/4" Globe A YIs alA Globe A A
A SP OP CL CL SP OP CL CL FC FC C,E,F,$,N,P,RM C,E,F,S,N,P,RM
((h)
- 03) -
[ 2 Out Out Yss 24'-4 3/4" Globe M M -
Yss 26'-6" Cate EM M OP OP CL Al YIs NA Cate EM E
E M OP OP CL Al B,G,RM B,G,RM 22g pggi,7 y Out Out YEs NA Check P P -
CL CL OP or CL -
Rev. Ffow 22 @,M IO b Id l l VIs 28'-3" Gate EM E M OP OP CL Al B,G,RM 22 1 Out Yrs NA Gate EM E M OP OP CL Al B,G,RM C
22 2 Out VIs 10'-6" Cate EM E M OP OP CL Al L,B,F,H,RM 15.5/15 1 In Yss NA Gate EM E M OP OP CL Al L , B, F , H,RM 15.5/15 2 In Yas 128-0" Cate EM E M OP OP CL Al B,G,RM Std. 1 Out Yss C NA Gate fM E M OP OP CL Al B,G,RM Std. 2 Out Yt's 148-0" Cate EM E M CL OP CL Al U,A,M,RM <33 1 Out Y2s 8 NA Gate EM E M CL OP CL Al U,A,M,RM <33 2 Out YIs NA Chk P P -
CL CL CL -
Rev.Floh - -
Out Yts NA Gate EM E M OP CL OP or CL Al J,F,K,M,T,RM ,V,0 20 2 Out Yes 13' 2" Gate EN E M OP CL OP or CL Al J,F,K,M,T,FM ,V,0 8
20 1 Out Yes NA Globe EM E M CL CL CL Al J,F,K,M,T,RM*,V,Q Std. 2 Out YJs 13' 0" Gate EM E M OP CL CL Al C,E,F,S,W,P,RM ,RM Yss NA Gate EM E M OP CL CL Al C,E,F,S,N,P,R'c M f
25 25 1
2 Out Out
)
J Yts 88-6" Gate EM E M CL CL CL Al L,B,F,H,RM 20/15(18) 1 Out YIs NA Gate EM E M CL CL CL Al L ,B, F , H , RM C
20/15(10} 2 Out Ycs NA Ret c P P -
CL CL OP or CL - - , , ,
l Revision 9 6.2-207 April,1998
, -- , n -
f- ~}1*i pag prcVIkg Q b c6 Attachment 4 TABLE 6.2-32 (Continued) PY-CEl/NRR-2352L Page 28 of 74 i
i NOTES: I
- 1. Through line leakage classification is discussed in Section 6.2.3.
l
- 2. Abbreviations used are as follows:
A - Air LPCS - Lov pressure core ADS - Automatic depressurization spray system system M - Manual i i AI - As is NA - Not applicable B' fly - Butterfly valve- OP - Open Chk - Check. valve j P - Process fluid i C1 - Closed RCIC - Reactor core E - Electric EH - Electrohydraulic isolation cooling system
- Electric motor Rel - Relief valve
[ EM FC H'
- Fail Closed RHR - Residual heat I
- Hydraulic removal system HPCS - High pressure core spray RVCU - Reactor water i system cleanup system
-LC l
- Locked or sealed in closed S - Solenoid !
position SLC - Standby Liquid I LPCI - Lov pressure coolant Control System injection system SP - Spring V - Vacuum in containment
- 3. Penetrations not listed are spares and are capped, except j penetrations P202 (Unit 1)/P306 (Unit 2) which are the equipment hatches.
- 4. Essential systems are engineered safety feature systems which are required for shutdown. In addition asterisk (*) items are non-ESF systems that vould be desirable to use to mitigate the consequence of'an accident. ,
- 5. Location inside (I) or outside (0) of containment.
- 6. Length of pipe from containment to outermost isolation valve.
- 7. All motor operated isolation valves remain in last position upon failure of valve power. All air operated valves close upon loss of motive air.
l l
l 6.2-212
. _ .. . ,. __ , - ~ .- , . _ . _ . _ . - . _
. - _. _ - _ . - _ . _ _ . . _ . - - .- - .. . _ . _ . - _ ~ . . - - --- - - - -
v -- ~-
T"L's. pc.y prt><.Aed- #
Or- .Jermb
^ on
^ - no cAag e.* Attachment 4 PY.CEl/NRR-2352L t Page 29 of 74 l TABLE 6.2-32-(Continued)
NOTES: (Continued)
- 8. Remote-manual (RM) valves can be opened or closed by remote-manual switch operation during any mode of reactor operation, except when an automatic signal is present. All remote-manual valves have position indicator lights at the remote-manual switch, and a subset of these valves have an additional set of position indicating lights at the control room isolation status panel.
Isolation signals are defined as follows:
Signal Description l A Reactor vessel low water level - Level 3. (A scram l / occurs at this level. This is the highest of the three isolation low water level signals.)
B Reactor vessel low water level - Level 2. (This is l the second of the three low water level signals.
l The reactor core isolation cooling and high pressure l core spray systems are activated at this level.)
C Reactor vessel low water level - Level 1. (This is
! the lowest of the three water level signals. Main steam line isolation occurs at this level. The low pressure core spray and low pressure coolant injection systems are also activated at this level.)
D Spare.
E Line break - main steam line (steam line high steam l flow).
F Line break - main steam line (main steam line tunnel high space ambient or high differential temperature).
G High drywell pressure.
H Line break in reactor water cleanup system (high ambient or high differential temperature).
( J Line break in steam line to reactor core isolation cooling residual heat removal system (low steam line pressure).
K Line break in reactor core isolation cooling system
- steam line to turbine (high steam flow).
l Revision 7 6.2-213 March, 1995 l
l l
l
y -
'M
$ ;3 png p rt, y i d &d Or' In O /' M N c m
" ^0 D6 Attachment 4 TABLE 6.2-32 (Continued) PY-CEl/NFR-2352L Page 30 of 74 NOTES: (Continued)
Signal Description L
High differential flow in the reactor water cleanup system.
i M Line break in residual heat removal system (high j ambient or high differential temperature).
! N Low main condenser vacuum.
P Low main steam line pressure at inlet to turbine (RUN mode, only).
Q Line break in reactor core isolation cooling system l
(high ambient). l S
High main steam line temperature, turbine building.
T High pressure reactor isolation cooling turbine exhaust diaphragm.
U- High reactor vessel pressure - close residual heat
/ removal - shutdown cooling valves and head cooling i valves. i'
! V I Line break in steam line to reactor core isolation l cooling residual heat removal system (high steam !
flow). I W High temperature at outlet of cleanup system nonregenerative heat exchanger.
X Containment to atmosphere differential pressure greater than 0.0 psid.
Y Standby liquid control system actuated.
Z High radiation, containment and drywell ventilation exhaust.
i RM Remote-manual switch from control room. (All A c automatically actuated containment isolation valves are capable of remote operation from the control room.)
I Revision 8 6.2-214 Oct. 1996 l
. ~ _ _ _ . . . . -. . = . _ - - . - . - . - - _ _ . . . . . - - . . -. __---
i
! Attachment 4 PY-CEl/NRR-2352L Page 31 of 74 TABLE 6.2-32 (Continued)
NOTES: (Continued)
Signal Description
- 9. Standard (Std) closure time, based upon nominal pipe diameter, is approximately 12 inches / minute for gate valves and approximately 4 inches / minute for globe valves. The standard closure time for butterfly valves is 30 to 60 seconds.
- 10. AC motor operated valves required for isolation functions are powered from the ac standby power buses. DC operated isolation valves are powered from the batteries.
- 11. Testable check valves are designed for remote opening with zero differential pressure across the valve seat. The valves close under reverse flew conditions, even if the test switch is positioned to open. The valves open when pump pressure exceeds reactor pressure, even if the test switch is positioned to close.
- 12. Deleted
- 13. Main steam line isolation valves require that both solenoid pilots be de-energized to close. Accumulator air pressure plus spring act to close valves when both pilots are de-energized. Voltage failure at only one pilot does not cause valve closure. These valves are designed to close fully in 2.5 to 5 seconds (see Section 5.4.5.3) .
- 14. During reactor operation, a blind flange is installed on the j outboard end of the transfer tube as the containment boundary.
- 15. Deleted
- 16. This receives a Type B test.
- 17. Deleted
- 18. Valve stroke times are specified as Unit 1/ Unit 2. (If only one time is indicated, it applies to both units.)
- 19. The ECCS response time requirement for these injection valves is met with the valves partially opened. Times shown in this table are for full (100%) valve strokes.
l I Z.0, b T_ u ur P
- 11. A 16 54.IY Revision 8 6.2-216 Oct. 1996
.. . ~ .. . . - _ _ _ . . . . -. . . - . - _ _ - _ .
AttIchmInt 4 PY-CEl/NRR-2352L Page 32 of 74 Insert Page 6.2-216
- 20. Division 3 power is also available by post accident operator manual action (procedurally controlled).
- 21. An Appendix J exemption for the Feedwater check valves documents that these valves utilize an alternate testing methodology involving visualinspection of the valve seats per the Inservice Testing Program to verify their proper closure.
1 i
o
TABLE 6.2-33 POTENTIAL SECONDARY CONTAINMENT BYPASS LEAKAGE PATHS ( ' (
Primary Containment / Line Bypass Expected m Penetration No. Size Leakage Air Leakage.
Unit 1 Unit 2 (3) I Description Valve No. Valve Type Loc. (in.) Barrier Rate (SCCM)
P106 RCIC Turbine E51-F0068 Gate 0 12.00 10(b) 190.0 2 Exhaust, RCIC E51-F0040 Noz.Chk. 10.00 0 10(b) 2.0 5 Turbine Exhaust !
Vacuum Relief, and E12-F102 Glb. 0 1.5 23.8 10(a)
RHR A&B Relief Valve Discharge to N27-F751 Glb. 0 1.0 37 0.08 l Suppression Pool P87-F083 G1b. 0 0.50 57 0.0 1 P87-F264 G1b. 0 0.50 57 0.0 P (15) i Y / I" U P108 P424 Condensate Supply Pll-F060 Btf. 0 12.00 T) 7
" 27(a) 0.63 i '
P11-F545 Chk. I 12.00 27(a) 384.48 f1 P109 P428 ILRT Blowdown Line (6) 't.
8.00 35 0.00 o P111 Y P426 Condensate Return Pil-F080 Btf. O 10.00 27(b) 0.52 -
I P11-F090 Etf. I 10.00 27(b) 0.52 7
P114 P121 Containment Vacuum M17-F015 Btf. 0 24.00 19 3.60 \,. o i Relief M17-F010 Chk. I 24.00 19 3.40 0-rings 24.00 19
/
I (13) 2 P115 See P106 0 0 P117 P413 Nitrogen Supply to P86-F002 G1b. 0 2.00 41 64.88 uu> l
{8m CRD P86-FS28 Chk. I 2.00 41 64.88 ggy i w m :r P119 P429 ILRT Pressure (6) - -
0.50 35 0.00 "E!
Indicating Line
}k
TABLE 6.2-33 (Continued)
NOTES:
- 1. A Technical Specification commitment of 5.04 percent of total design containment leakage is made for l the maximum test leakage bypassing the containment annulus exhaust gas treatment system (for leakage sources in lines penetrating primary contairment and personnel airlocks).
- 2. Expected bypass vater leakage sources from components that circulate core cooling vater following LOCA:
i cw
- a. Pumps - total leakage of S gal./ day from residual heat removal, high pressure core spray, and low pressure core spray pumps.
tw 4
- b. Valves oa valve stem leakage of 300 cc/hr from systems handling reactor fluid outside m containment.
L L c. FV-LCS, ECCS and RCIC Branch Lines - total leakage of 1.663 gal./hr.
U
- 3. Bypass leakage barrier arrangeaent is shown on the designoted detail of Figure 6.2-60.
- 4. Closed Systems Outside Containment
. u \;n~N ca %gusw&ar % % gls g g_
Piping systems which penetrate containment and are closed outside containment are tested for potential water leakage under the guidance of NUREG-0737 tem III.D.1.1. The expected bypass water l leakage from these systems is identified in Note 2 above. Valves in closed systems outside containment, which are potential air leakage sources, are identified in the text of Table 6.2-33.
The redundant containment isolation provisions for each penetration consist of an isolation valve and xx a closed system outside containment which are in compliance with 10 CFR 50, Appendix A, Criteria 54. yTR
$$ The closed system is mis $de protected, Seismic Category I, Safety Class 2, and has a temperature and Wof
@.7 pressure rating in excess of that for containment.
Wgi E Emi 5" - 5$* l a i y
i
TABLE 6.2-33 (Continued)
NOTES: (Continued)
The following penetrations lead to closed systems outside containment:
Unit 1: P101, P102, P103, P104, P105, P107, P112, P113, P118, 123, P132, P401, P402, P403, P407, P408, P409, P410, P411, P412,LP421, P429, P431.
Unit 2:
L
, Yl2.lh P101, P102, P103, P107, P106, P108, P109, P113, P114, P111, P117, P133, P408, P401, P402, P403, P404, P405, P407, P409, P411, P417, P418, P406, P419, P421.
- 5. h Liutrt ~Ins tt wuent Lines 3
Instrument lines penetrating containment are assumed to allow zero bypass leakage. Valves in instrument lines penetrating containment are open post-LOCA in order to fulfill the instruments'
, functions. The instruments are designed to allow zero bypass leakage. Penetrations containing g instrument lines appear below:
Unit 1: P102, P318, P320, P401, P402, P425, P433, P434 Unit 2: P102, P422, P423, P401, P402, P219, P220, P221 4
- 6. Spectacle or Blind Flanges Penetrations which contain lines isolated by spectacle or blind flanges are assumed to allow zero bypass leakage. Spectacle and blind flanges are type C tested. Any leakage determined by type C tests will be eliminated by tightening and/or re-sealing the flange. Penetrations whose lines are isolated by spectacle or blind flanges appear below:
mm>
oy Unit 1: P109, P119, P120, P205, P317, P319 yQ ct <
w hk Unit 2: P428, P429, P427, P304 { jam [7
=8
- e.
m r$
M
Attichment 4 PY-CEl/NRR 2352L Page 36 of 74 Insert Page 6.2-226
- These feedwater penetrations have a branch line leading to the RHR Shutdown
. Cooling Return Line and the RWCU Return Line. The RHR Shutdown Cooling Return Line leads to RHR, which is considered a closed system outside containment. The RWCU Return Line leads back into the containment, and is considered a closed system outside containment, with the specific acceptance criteria that leakage exterior to the piping will be eliminated (see Table 6.2-40 for testing details on containment ,
l j
l
)
l i
l
. . _ _ . . ~ . - . . , ~ . - -
TABLE 6.2-33 (Continued)
NOTES: (Continued)
- 7. Leakage Control Systems The main steam lines have been excluded as potential bypass leakage paths. The main steam isolation valve leakage control system (refer to Section 6.7) controls leakage from the isolation valves. The main steam lines go through the following penetrations:
1 Unit 1: P122, P124, P415, P416 I Unit 2: Pil5, P116, P415, P414 3~'CS'J r~
The feedwater system has a dedicated leakage control system (refer to Section 6.9) which pre::urie :
th; piping betu :: th: inheard th:th valv:: and cuth: rd ;;t: 7:17c. Onli th cirbczn: fr;; tion of
, any ft:6::ter lethage frr- the F -
LCS -ill bc :::idered :: byp :: 1ech:gc. Th r" LCC ::al: lin :
, ga thrcugh the fellering penetratiene:
E o
w Unit 1: P121, P414 Unit 2: Pil2, P410 The leakage control systems meet single failure criteria, are missile protected, Seismic Category I, Safety Class 2, and have temperature and pressure ratings in excess of that for the containment.
8.
The personnel airlock door seals are not considered a bypass leakage path when the outer door is operable because all leakage through the outer door seals is routed back into the annulus area between the annulus shield wall and containment where it is treated by the Annulus Exhaust Gas Treatment System.
o Om
- o M 4
~ ~.
W TT) aog
$O m
m N za3 c
"Ed Nw A r"-
Attachment 4 PY.CEl/NRR-2352L Insert Page 6.2-227
.provides seal water to the bonnet, stem, and seat of the outboard gate valves (B21-F065A/B). Water leakage from the Feedwater Leakage Control System (FWLCS) piping, and from the bonnet, stem, and seat of the outboard gate valves, are controlled under the Primary Coolant Sources Outside Containment Program, Technical Specification 5.5.2. Outboard gate valve bonnet and stem leakage identified during the -
system walkdown at pressures >1000 psig will be eliminated. The gate valves will be closed during this check. Gate valve seat water leakage measured at 21.1 P. will be limited by the Program, which restricts allowable leakages to half of that assumed in the dose calculations (see Section 15.6.5.5.1.2.b). This water leakage is not added into the secondary containment bypass air leakage totals. The FWLCS seals the feedwater lines going through the following penetrations:
F
4 TABLE 6.2-40 y
PRIMARY REACIDR CONTAIPNENT PENE'rRATION AND COffrAl?NENT ISOIATION VALVE LEAKAGE RATE TESf I,IST 4
Inboard Outboard Containment contalment Fenetration Pene- Inboard Isolation Barrier Cutboard No. Utilt 1/ tration Barrier Barrier Description / Isolation Barrier !
7 Unit 2 Description Barrier Barrier Description / 3 Test Test Valve No. Notes Test Valve No. Notesk$,
P202/P306 Equipment hatch B Ibuble 0-ring 1 - -
P305/P205 tower Personnel Airlock Barrel B Inner door 2 Inner door Inflatable gaskets 1 Outer door Inflatable gaskets 2
1 g
C P53F536*/F570 3 D j C P53F015*/F070 -
C P53F556 C P53F035 7 ; 3* j1 C
C M F160 P53F010*
gp t
i m C
' P53F030 -
l ,3 u, >V 1 M
U (* inboard / outboard function varies with mode of operation) i h
P312/P215 Upper Personnel Inner door 2 outer door 2 Airlock Barrel B Inflatable gaskets 1 Inner door 1
C Inflatable gaskets 1 fg P53F541*/F571 3 jZ C P53F025*/F075 -
e C P53F561 C P53F045 7 j C P52F170 -
C MIF020* -
I C P53F040 - i r
(* inboard / outboard function varies with mode of operation)
?5 cv < P205/P304 Fuel transfer tube B Double gasket 1,12 -
TT>
g-se. P124/P116 Main steam line A B C B21F022A 3,4,12,13 C af?
$@ B21F028A 4.13 *MS m C C
B21F067A E32F001A 13 13
$gg Rya 5?* o Un 6
- - - _ . - - _ _ . _ _ _ _ _ . - _ - _ _ - - - . - _ - - _ - - - _ - _ _ _ _ _ _ _ _ _ - - - - _ .--. - --.._ - _. , - . _ . + -- c -
f i
TABLE 6.2-40 (continued) !
I 4
Inboard Outboard Containment Containment. ;
Penetration Pene- Inboard Isolation Barrier Outboard Isolation Barrier '
tb. Unit 1/ tration Barrier Barrier D=scription/ Barrier Barrier Description / .
Unit 2 Description Test Test Valve tb. tbtes Test Valve No. tbtes. {
P416/P414 Main stea:n line B B C B21F022B 3,4,12,13 C B21F02BB 4,13 ;
C B21F067B 13 i
- C E32F001E- 13 i P122/P115 timin stea
- n line C B C B21F022C 3,4,12.13 C B21F028C 4.13 ,
C B21F067C 13 ;
C E32F001J 13
[
0415/P415 Msin steam line D B C B21F022D 3,4,12,13 C B21F028D 4,13 ;
C B21F067D^ 13 -
C E32F001N 13 i i
m P121/P112 Feedwater A, PRR, B / N27F559A 5, 13 /= B21F032A 5,13 I RWCU Return to y B21F065A 5,13 I Y Reactor Pressure c Ett FoSA A 7,12 1
$ Vessel --
b)CSC,l M Qtm 13,30 l P414/P410 Feedwater B, PRR, RWCU Return to B
g N27F559B 5, 3 hC B21F032B B21F065B 5,13 5,13 j
Reactor Pressure c Et1 fos 3B y, g3 i vessel .
y ctg cl.wA Syg+aa t3,30 [
P102/P102 RHR punp A suction -
C E12F004A 9,13,17 -
Closed system 7 :
P402/P402 RHR purp B suction -
C E12F004B 9,13.17 -
Closed system 7 f
i P403/P403 RHR purp C suction -
C E12F105 9,13,17 -
Closed system 7
[
TT> t
@y P421/P406 RHR shutdown cooling B C E12F009 12,18 C E12F008 18 $ $
et < suction C E12F550 18
- w- 1 rn S t 4
o-3 ;
M -
$3
? -
EWa Y
t
! ?
l 9
i
, 1
N
. TABLE 6.2-40 (Continued) !
Inboard outboard containment Containment 4 Penetration Pene- Inboard Isolation Barrier Outboard Isolation Barrier .
?
No. Unit 1/ tration Barrier Barrier Description / Barrier Barrier Description /
Unit 2' Description Test Test Valve No. Notes Test Valve No. Notes P410/P411 HPCS punp discharge -
C E22F005 13,25 C E22F004 13,25 i to RPV Closed system 7 -<
, P409/P409 HPCS min. flow and --
C E22F012 13,16- -
test line to C E22F035 8,13,16 -
Closed system 7 A suppression pool C E22F023 13,16 Closed system 7
{
2 P103/P103 LPCS punp suction -
C E21F001 9,13,17 C Closed system 7 l
P112/P113 LPCS pungd' ischarge -
C E21F006 13,25 C E21F005 13,25 .
, to RPV >
P423/P423 Main steam line drain B C .B21F016 3,12 C B21F019 -
4
[t 3
C P131/P132 RWCU punp suction B C G33F001 12,13 C' G33F004 13 or P419/P432 RWCU punp discharge -
C G33F053 -
C G33FD54 -
- p P132/P408 RWCU return to- -
C G33F040 -
C G33F039 -
0[
4 feedwater I)
P424/P420 RWCU to unin -
C G33F028 -
C G33F034 -
lId condenser and radwaste C G33F0646 8
{7l o l P203/P301 Fuel pool cooling -
C G41F522 - C G41F100 - '
end cleanup supply p sw 4>>r,,w U. S. $[E.
[E
.. @ reE!!.
- 3 a
r
't 9
___.___.__________________.___.m_._.____ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ . _ _ _ _ _ _ _ . . . . _ _ _ _ _ _ _ _ , _ _ _ _ _ , _ _ _ _ _ _ _ _ . _ .-- , ._e .--
. . - - . - - - - -_ . . - _ - - .- ~ - - - . _ _ - . - _ _ - ._ .
Attachm:nt 4 P Y-CEl/NRR-2352L Page 42 of 74 TABLE 6.2-40 (Continued)
NOTES: (Continued) n b
5.
e feedwater lines ~are sealed post-LOCA with water from the feedwater leakage control system (FWLCS).
Feedwater lines wil tested as follows. The inboard and outboard check valves 1 be tested with water to a pressure not less than 1.10 P . Acceptable check valve leakage is 1 gpm per valve. The ou ard gate valve stems and bonnets will be Type 'C' tested h air. A high
- pressure (1,000psig)waterleaktes f the outboard gate valve f stem and bonnet may be perfome the Type 'C' air test, wit s an alternative test in lieu of zero water leakage being acceptable.
Water leakage throug e check valves is not included in the 0.60 LaType B C test totals. Also, if a Type 'C' test with air is per
.ed, air leakage through the gate valve stem and bonn will be eliminated. Gate valve through-seat leakage is not nsidered bypass leakage.
- 6. System remains water filled post-LOCA.
Isolation valve tested with water to a pressure not less than 1.10 P . Isolation valve leakage a
not included in 0.60 L, Type B and C test totals.
kur Y 7.
The redunda t containment isolation provisions for this penetration /
consist of f inflatable seal and an isolation valve. A si active failure can be accomodated. For the leak ontrol function, which is applicable to the s Ine from outer door to annulus, the line to the annu s designed to standards for a closed system to se ry containment. The closed system is missile ed, Seismic Category I, Safety Class 2, and has a erature and pressure rating in excess of that for the 6.2-250 Revision 8 Oct. 1996
Attachment 4 PY-CEl/NRR-2352L Page 43 of 74 Insert for NOTE 5 Page 6.2-250 The Appendix J Type "C" test for the Feedwater lines is provided by the Type "C" hydrostatic tests performed on the long term, high integrity leakage protection valves, i.e., the motor operated gate valves (821-F065A/B). Water leakage from the Feedwater Leakage Control System (FWLCS) piping, and from the bonnet, stem and seat of the motor operated gate valves, are controlled under the Primary Coolant Sources Outside Containment Program, Technical Specification 5.5.2. Outboard gate valve bonnet and stem leakage identified during the system walkdown at pressures >1000 psig will be eliminated. The gate valves will be closed during this check. Gate valve seat water leakage measured at 21.1 P will be limited by the Program, which restricts allowable leakages to half of that assumed in the dose calculations (see Section 15.6.5.5.1.2.b).
This water leakage is not redundantly added into the Type C 0.60 L. totals, secondary containment bypass air leakage totals or the hydrostatic test program totals. An Appendix J exemption documents that the Feedwater check valves (N27-F559 A/B and B21-F032 A/B) utilize an alternate testing methodology involving visualinspection of the valve seats per the Inservice Testing Program, to verify proper closure.
Insert for NOTE 7 Note to reviewers:
The following insert for NOTE 7 on page 6.2-250 is simply restoring the correct wording for this NOTE regarding closed systems outside containment. A change made in i Revision 8 of the USAR had incorrectly revised NOTE 7 such that it became a specific note for just one type of penetration (the airlocks), versus the generic note that it had historically been. This is being corrected under the PNPP Corrective Action Program.
)
i The insert to NOTE 7 therefore, is not a change being made as a result of the proposed feedwater penetration improvement amendment, per se, other than the addition of the l reference to GDC 55.
...an isolation valve and a closed system outside containment which is in compliance with 10 CFR 50, Appendix A, GDC 54 and 55. A single active failure can be l accomodated. The closed system is missile protected, Seismic Category 1, Safety Class 2, and has a temperature and pressure rating in excess of that for the containment. Closed system integrity is maintained and verified during periodic Type A test and during system leak tests (per NUREG-0737 Item Ill.D.1.1).
l
& s,' g gf- M br w hom. No (O f f"'f f _
-v - w TABLE 6.2-40 (Continued) Attachment 4 PY-CEl/NRR-2352L Page 44 of 74 NOTES: (Continued) 12.
Penetration design utilizes a double bellows for containment isolation. The bellows is leak checked by pressurizing the space between the inner and outer bellows. The fuel transfer tube bellows is sealed on both ends with double gasketed flange joints.
These joints are leak checkec'. by pressurizing the space between the I double gaskets.
The fuel tra.nsfer tube has a bellows assembly installed in the annulus to p'trmit confirmatory leak testing of the fuel transfer tube bellows. The fuel transfer tube bellows is leak
' checked by pressurizing the annular space between the IFTS tube and the bellcws assembly via a test connection on the bellows assembly,
{
1
- p. 13.
System is not vented and drained for Type A test. This allowance for main steam lines is provided by Reference 33.
14.
Isolation valving for instrument lines which penetrate the containment confonn to the requirements of Regulatory Guide 1.11.
The ISI program will provide assurance of the operability and integrity of the isclation provisions. Type 'C' testing will not be performed on the instrument line isolation valves. The instrument lines will be within the boundaries of the Type 'A' I
test, open to the media (containment atmosphere or suppression pool water) to which they will be exposed under postulated accident conditions.
Three exceptions to the above are Penetrations P401, P318 and P425.
Isolation valves for the three penetrations include H
2 Analyzer and Postaccident Sampling System Valves. The valves are normally closed post-LOCA, open only intermittently, and will~
therefore receive Type C tests.
- 15. System remains pressurized with air post-LOCA.
Revision 8 6.2-252 Oct. 1996
- - . - . . - - . . - . . .. . - . ~ . . - - . . _ . - . . - . . - . - . . . - . . . . - . - . . . - - . _ _ . . . . . . - . . . . -
Attachm9nt 4 PY-CEl/NRR-2352L-Page 45 of 74 2 8 .~ The containment purge 18-inch supply and exhaust lines are provided with double inboard isolation valves. For each 18-inch line, both inboard valves are Type C tested and the highest leakage is designated as the inboard barrier leakage. Leakage through the test. connection between the 18-inch isolation valves is summed with the innermost 18-inch valve leakage.
- 29. Deleted 30 In W Y J
l l
2 Revision 8
- 6.2-255a Oct. 1996 l l
l - - , - , - - -_.
.____..m _ . _ _ __ _ _. _ _ _ _ _ . _ _ _ _ . _ _ . _ _ __
Attachment 4 PY-CEl/NRR-2352L i
Page 46 of 74 Insert Page 6.2-255a
- 30. This RWCU line returns the filtered RWCU water to the Reactor Vessel via the l Feedwater lines. The piping " outboard" of the RWCU branch line check valve l (1G33-F052A/B) leads directly back to containment penetration P132, and is j ASME Code Class 2, Seismic Category 1, protected from pipe whip, missiles and l jet forces, and analyzed for " break exclusion". This closed system outside containment contains only mechanicaljoints, including the packing on the outboard containment isolation valve (G33-F039) for Penetration P132 (see the l P132 entry). This outboard valve (G33-F039), including the stem and bonnet, is already part of the air leak rate test program. The remainder of the RWCU piping
- between the Feedwater line and Penetration P132 will be added to the Technical l Specification 5.5.2 Primary Coolant Sources Outside Containment Program, with a specific leakage acceptance limit of zero (0) water leakage when tested at
> 1000 psig.
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@ PERRYNUCLEARPOWERPLANT
$uwea cow Containment and DrywellIsolation Figure 6.2-60 (Sheet 1 of 4)
(Dwg. D-300-76l) 0 0 ND l 2 - .
-- f his f' protddEd _ _-
fr#~ id'~hb o s " NC 6.9 FEEDVATER LEAKAGE CONTROL SYSTEM Attachment <4 PY-CEl/NRR-2352L Page 48 of 74 6.9.1 DESIGN BASES
- a. . The feedvater leakage control system (FVLC) is designed in accordance with Seismic Category I and quality group classification requirements to comply with Regulatory Guides 1.26 and 1.29. The system meets the intent of Regulatory Guide 1.96, where applicable (see Table 3.2-1).
- b. The FVLC system is designed with sufficient redundancy, separation, reliability, and capacity as a safety-related system consistent with the need to maintain containment integrity for as long as postulated LOCA conditions require.
- c. The FVLC system is capable of performing its intended safety function following a loss of all offsite power coincident with the postulated design basis LOCA.
- d. The FVLC system is designed with sufficient capacity and capability to prevent leakage through the feedvater lines consistent with containment integrity under the conditions associated with the postulated design basis LOCA.
- e. The FVLC system is provided with interlocks actuated from appropriately designed safety systems or circuits to prevent inadvertent system operation.
- f. The FVLC system is designed to permit testing of the operability of I controls and actuating devices as well as the complete functioning of the system during plant shutdowns.
I 4
l 6.9-1 r - c mr 4 --- w
l Attichm:nt 4 PY-CEl/NRR-2352L Page 49 of 74
- g. The FVLC system is designed so that effects resulting from a system single active component failure vill not affect the integrity of the feedvater lines or the operability of containment isolation valves
- h. The FVLC system is protected from the effects of internally generated missiles, pipe break failures and adverse environments associated with a LOCA.
L. Lse.r t 6.9.2 SYSTEM DESCRIPTION The FVLC consists of piping, valves and instrumentation as shown in Figure 6.9-1. The system components are designed to the requirements of Table 3.2-1, Item XLIX.
The FVLC system consists of two drde;crd:n: subsystems designed to eliminate throuch line lenkare in the feedvater oloing by providing a (fer h 5 *^A 5 a t c+ % eJ positive sealXo:: :: tw, bbanot
. . . . n: ; ;n- t - ' - t i c n C. : d o1 ~r :-d h Cea e a h na outboard isolation valv (bM sim O he W ubs stem uses the residual heat vision 1_)
removal (RHR) vaterleg pump and the a :Lem u i uosystem uses the Inu - 9 'y pressure core spray (LPCS) vaterleg pump to supply sealing wat(C&. g : ..:
d:an:::::: :nd :p:::::: :!d:: cf th: :::b::rd ::rtedr er: 1 :12:!:r ch::5 ::1::, :::;:::i.cly.
Following a LOCA, the FVLC system is manually initiated from the control room. The operator first verifies feedvater unavailability through low feedvater pressure (approximately 30 psig), then closes the outboard containment isolati tor operated gate) valves with the keylock ess o switches, and open ie motor operated FVLC system valvec f*e= t e eo< A %
I control room. The su on pool sealing water fro, waterleg moVs.
pumps is routed to bot p) rat:r li=0. The :aling f!u!" frer 5: """
--: rle; pu ; dir 9 rg !! f!11: 22:5 fort t - 14ne 6 t t'a nn tha
- cr* !- n* !:21 t! :5 P valver. T5: :al! .; !cr " ugh the :1 :
eve ll) fill: th: f:: dst r li:w up ' th: ::::::: ::::1 (or 1 ::
(or bo% M ney are. M all
- M 6.9-2 O
ad- - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - .
l Attachment 4 PY-CEl/NRR-2352L Page 50 of 74 Insert Page 6.9-2 l
- l. The normal power supply to the Feedwater penetration motor-operated gate valves is from Division 1. The licensing basis for the feedwater penetrations is i
that the gate valves are successfully closed by the control room operator. Power is also available to these gate valves from Division 3 (altemate pov.er source connected per plant emergency procedures). This improves the reliability of the penetration in the event of a total loss of both the normal and emergency AC power from Division 1. Physical and electrical separation between Divisions i l
and 3 will be rnaintained during normal operation by employing two features:
- 1. Normally open, fused disconnect switches at both ends of the circuit, and
- 2. Fuses normally stored out of the circuit.
Insert
...through the bonnets of the MOVs. Following closure of the MOVs, the sea!ing water seals the stems, bonnets and seats, and isolates the feedwater lines.
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Attachment 4 PY-CEl/NRR.2352L Page 51 of 74 th; dryr:11 in the r2rr rfafert2t rpip-break inride drfr:11' 1rd finni!y th: r:ter returrr tr *' rrpp----i^" p^^' "'-cu;F t 's LOC?_
hrer' Since the source of sealing water is the suppression pool, a 30-day water supply is ensured. Operation of the FWLC system will not affect the function of the suppression pool since the :- ' rter i "- --
eventually return the pool when the dry W flooded back over the weir wall. ahM v obe It 4 *t5 ; &6 ofhC"MAM is,very s % h j u&. .ag b mNaw4 mQ Th. :;;1ing unter frer *'- LPCF "-ter! ; p" r dir- ge if-^ fill: :::h
-I.;d ;;;r lin; h:tu:: th: uth: rd ::ntrir nt i :12ti:n :h::h valc;
- nf th: f::fr ter ch t:f f ;2tr vil- c.
When the FWLC system is initiated manually following a LOCA, there should be no demand for keep-fill water in the RER and LPCS systems since these systems will be operating. Therefore, the waterleg pump should be totally dedicated to provide sealing water to the FWLC system. l A single waterleg pump has the capacity to provide the necessary sealing water to the FWLC system.
The feedwater system will not be completely drained since the system will be intact and operating initially post-LOCA.
The feedwater design includes a backup flow path through a motor driven pump. When the turbine driven feed pumps lose driving eteam and trip on vessel Level 2 post-LOCA, flow is automatically diverted through the motor driven pump. The motor driven feed pump and/or the feedwater booster pumps will continue to pump water into containment post-LOCA.
The pumps will continue to operate for about 10 minutes before the feedwater booster pumps trip on low water level. During this time, no extraction heating is available and cold water from the condenser hotwell is being pumped into the vessel which cools down the feedwater Revision 9 6.9-3 April, 1998
Attachment 4 PY-CEl/NRR 2352L Page 52 of 74 and the piping. When feedvater flov is finally stopped, feedvater flashing is not expected to occur. Therefore, a significant voiding of the piping is not expected.
is case of a LOCA where the feedvater lines remain water filled, static hea ssure and the feedvater check valve disc veight force the feedvater check valve close. Initiation of the FVLC system pressurizes the volumes between inboard and outboard isolation check valves, and between the outboard isolation valve and the motor operated gate valve. Since the waterleg pumpc operat .
pressure higher than the static head pressure, leakage is into the reacto vessel (or dryvell as discussed previously).
In the case where the feedvater lines do not remain completely water filled, the feedvater system can be operated to ensure positive pressurization up to the motor e A
(it & coerated i.oc A h sente Fur n>ar valve. ur_m and sur- thusu>% leakare & 1 [J '
ir'tothereactorvessel(ordryveHp: d i : C r '. : ;'::.i c;17). 2ne t wu; l system vill be initiated and begin to fill the volume @t"te" t![h 6.d]
!-d::: d _::d : _ t b- _: d ! n!: t ! n :!: v el n , 2::' E:' ::: ' ' - m iE yA ht of I h Mor 3_ 7_.3_ _t
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": :1: . . . When the volumes fill up, the pressure vill increc O q q
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- very little voiding is expected, a conservative approach to l
evaluate the FVLC system wuuu 'n 'n nssume one feedvater line is completely drained. Fill time in this case vould be appiur.... tQ b 18 minutes. fIf a divisional failure is assumed, -!y :: waterleg pump
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i l 6.9-4 1
I
Attrchmsnt 4 PY-CEl/NRR-2352L
. o t s $4.A \ o y Page 53 of 74 pil prve y^3 a wouldgbe available for filling the feedvater line. Under these conditions, calculations sho' would take -p; r- : al, % minutes to fill and maintain a feedvate e$1. k8 8 M" 9 GperwimstsAQ G or 6eWot)
Basedontheconservativeassumpionslistdabove, the FVLC cystem vill provide an adequate seal withindone hour follovi a LOCA. If a loss of offsite power is assumed at this time, the FULC tehvillmaintainthe Qn %.- eam i 4 a: a:::M a-t an A w.At
- = ;cn n 9 "- h e 2nd th;-
volume of watetAL ;.
(Tus Ac<t) outboard motor operated gate -valves, During :hi; ; .qhour, operation of the feedvater system vill maintain a system pressure higher than the dryvell pressure, thus ensuring water leakage into the vessel.
ks nadaining A f.eaaW 6.9.3 DESIGN EVALUATION Qw is.ol att on.
The FULC system is designed to prevent the release of radioactivity -m s tem %\ Deer ss.s.\ ,) ~
through the feedvater line isolation valves by provi ng ::nunuou:
fl:_ cf ;:ter Fre>f +be feet'eter l Mer following a loss of all offsite power coincident with the postulated design basis loss-of-ennlant accident. The otun w m;nm wn~+
redun subs u vex Wlantocsex)ystems are physically separate to minimize the exposure 01 the system components to missiles and to the ef fects of pipe whip or jet impingement from high energy line
- breaks. N ' * ' * ' ^ * ^ Pi f; %wt 3 mrs- ekyshAt pc % tea %.
~(% poektsA*.k]off a.-ekyS The FVLC system is Seismic Category I and is capable of performing its intended function following an active component failure. Each I i, W :n krr subsystem is powered from a different division of the ESF power supp11. Th poss.hi tip efr a shle- ac%* Nbesdons d hh Fu.Ap, A.+ar y 44ew va b - o g a.fe d vad veA- (.wbh Ar. a phet cf %
F=A ws+u- s,sh W 4% y F4WA+u- L AWY- b.nM( L MaQ i 3 Double series isolation valves are providedgto ensure that no singlel g active failure vill affect the integrity of the feedvater lines. b Mk a u pwM O -I ./.
l L supp ty U ceJ.3 1 6.9-5 l
._ m _ _ . _ _ . . _ _ . _ _ _ _ _ _ _ . . _ . _ - _ _ . . _ _ _ _ . . . . . . . _ _ _ . . _ . . . . . . _ _ . . _ . - _ _ . _ _ . _ _ . _ . . _ _ . _ . _ _ .
f Attachment 4 PY-CEl/NRR-2352L Page 54 of 74 .
Insert Page 6.9-5 in the event that the feedwater system becomes inoperable during the rapid vessel
- depressurization following a LOCA, the water in the feedwater piping will begin to flash into the drywell. It is expected that a water seal would remain for a sufficient length of time fo' lowing the accident until the operator remotely isolates the motor operated valve.
- Thus, a water seal would exist in the piping beyond (outboard)of the motor operated
- valve, initiation of the FWLCS to the bonnet, stem and seats of the motor operated valve will then provide the water seat for the remainder of the 30 days.
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_. . - . - _ . . . . _ . . . . . - . . . _ . ~ . . . _ _ ~ . . . . . . _ . _ . . _ _ _ . - . _ _ _ _ . . . . . _ . . - _ . - _ . _ _
Attichm nt4 PY-CEl/NRR-2352L Page 55 of 74 Non-seismic systems and components in the area of the FULC system have !
been analyzed for the effects of their failure. Additional supports or- l
. protection by barriers is provided to assure the FVLC system is not '
jeopardized by non-seismic failures during an earthquake. A single ,
failure analysis of the FVLC system is contained in Table 6.9-1.
6.9.4 TESTS AND INSPECTIONS i The FVLC system is hydrostatically tested prior to startup. The complete functioning of the system. including operability of controls and actuating devices, can be tested during periods of plant shutdown, but in no case at intervals greater than two years. '
i 6.9.5 INSTRUMENTATION REQUIREMENTS ;
Each FVLC subsystem is manually initiated from the control room !
following a postulated LOCA. Independent pressure instrumentation is !
provided for each FVLC subsystem in order to prevent operation while the feedvaterlinesarepressurizedjThemotoroperatedvalveon utboard subsystem of the FVLC system is interlo th each feedvater shutoff valve to preclude initia t e subsystem of the FVLC system g when th W M vaAve is not fully closed.
) Qyr e l l Div i Sion 1 4 A F W Lc. s3sha j is i d u-1oc.t.aa. % pukAt iM h Ho^
I sus A uhc o par a +d s L+c{+ ,
l f V A \ve,s wrv c,los 4 E cpar W cn o f- bi d s ; ..s Z-is mb ha A $
p6t peu / '.ns wcH o ss.
1 i
6.9-6 L
\
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. . , - . . ~ . . . . . - . . ~ . . _ - ..... _ .
l l
Attachm:nt 4 l l
PY CEl/NRR-2352L l Page 56 of 74 i
TABLE 6.9-1 1 SINGLE FAILURE ANALYSIS OF FEEDVATER LEAKAGE CONTROL SYSTEM i
COMPONEN'I /EQUIPMLNT MALFUNCTION CONSEQUENCES RHR vaterleg pump Either pump fails to One subsystem is LPCS vaterleg pump operate inoperative. System requirements met by redundant pump and associated subsystem. !
Hotor operated valve on Either valve fails to One subsystem is either waterleg pump open inoperative. System i discharge line requirements met by Q
l redundant subsystem.
Feedvater shutoff Feedvater shutoff valve e outboard subsystem valve fails to close as ciated with the fail valve remains deacti ted by virtue of an in rlock signal from shuto- valve position sv h. The FULC system requirements ar met by the redundant nboard subsystem. !
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l
Attichment 4 PY-CEl/NRR-2352L Page 57 of 74 Insert Page 6.9-7 Division 1 AC electrical Totalloss of both the Division 3 electrical power power normal and emergency is made available to the Division 1 AC electrical Feedwater system motor power sources. operated gate valves by manual operator action per plant emergency procedure. The Division 2 FWLC subsystem is then utilized to provide the water seal on the MOV.
l Insert to the "Feedwater shutoff valve malfunction" ite.n The Feedwater Leakage l Control System is l vulnerable to this extremely unlikely event. However, -
l the licensing basis for the i Feedwater penetrations is '
that a water sealin the l 1
Feedwater piping outboard of the shutoff valves would j remain for a sufficient i length of time following the i accident until the control I room operator successfully isolates the motor-operated valves. Therefore, these valves are assumed to work. Also, each Feedwater line contains two check valves which are classified as containment isolation valves on this penetration, and they provide a redundant containment isolation function. The proper closure of the check valves is verified by a visual inspection performed per i the Inservice Testing '
Program.
1 I
Attichm:nt 4 PY.CEI/NRR-2352L Page SB of 74 1
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- 5. oVRING INJECil0N TO THE BONNET Or VALVES 1821-r065A/B 10 FORM A PRESSURE SEAL. PRES $URE REACHES A uArluvu Or 39 PSIC A1 PT-NS98A/S. rLOwRATE AT OPERATINC DATA PolN1 1 DECREASES TRou 18 cpu 10 0 Cou. AND rLowRATE AT OPERATINC '
DATA POINT 2 DECREASES FRou 9 cpu 10 OGPM. SEE DCC-003 10 CALCULAll0N N27-20. REV. O AND CALCULAtl0N N27-45 REV. 2 1
l j
g PERRY NUCLE AR POWER PL ANT IPNob) i THE CLEVELAND ELEC7RIC b ILLUMINATING COMPANY l
! Feedwater Leakage Control System Figure 6.9-1 (Sheet 1 of 2)
(Dwg. D-302-971) i
AttachmInt 4 v
PY-CEl/NRR-2352L Pags 59 of 74 g S P,W f n, y l ,L, d for i n fo r m H on - Ao
_ A J dyWs /
i l
available to permit simultaneous safe shutdown of both units under all conditions with only one startup transformer in service, as discussed in Section 8.2.1.2.2.
The normal preferred source of power to Class 1E equipment is from the l unit startup transformer through the 13.8 kV startup bus and one winding of the 13.8/4.16 kV, two winding secondary, interbus transformer. One unit interbus transformer secondary winding feeds the 4.16 kV Class 1E load of the associated unit. The other secondary winding of the interbus transformer feeds the 4.16 kV Class 1E load of the other unit I through a normally closed circuit breaker. In addition, the Class 1E system of each unit can be fed from the startup transformer associated with the other unit or from it's own unit auxiliary transformer (Unit 2 i auxiliary transformer is future). All power supply selections are accomplished manually from either the control room or from a remote location. Both the startup transformers and the interbus transformers are sized to supply power to the associated unit Class 1E buses under l
LOCA conditions, and to supply power to the other unit Class 1E buses !
for use in safely shutting down that unit. The startup transformer and interbus transformer impedances were selected with due consideration to the fault duty of the breakers and the voltage regulation on the Class 1E buses. Calculations established the need for the present MVA ratings on the 13.8 kV and 4.16 kV switchgear. Calculations indicate the starting of the largest motor (6,000 hp), presents a relatively small voltage drop (less than 5%) which is considered insignificant.
8.3.1.1.2 Class 1E Power System The Class 1E power system is illustrated on Figures 8.3-1 and 8.3-2.
The system is designed with independent divisions having radial systems through all voltages at 4.16 kV and below. Complete physical and electrical separation is maintained to ensure maximum integrity. Note, Division 3 is capable of being manually cross-tied to Division 2 during Revision 7 8.3-2 March, 1995
m.. . .-_ - . .. .. . _ __ . . . . _ . - - - . = _ , . - . - - - - ----. . . _ _ . _ . . . . _ - - . - - _ . _
Attechment 4 PY-CEl/NRR 2352L '
Page 60 of 74 a station blackout to provide power to some Division 2 loads. Refer to appendix 15H for.more information. Switchgear associated with each '
division is housed in rooms within the
.1^ .tt.t- D W 9 s
h Revision 7 8.3-2a March, 1995
Attachmtnt 4 PY-CEl/NRR-2352L Page 61 of 74 Insert Page 8.3-2a Additionally, Division 3 is capable of being manually connected to Division 1 following a loss of coolant accident and a total loss of both the normal and emergency Division 1 AC electrical power sources, to provide power to the motor operated gate valves in the Feedwater lines. Physical and electrical separation between Divisions 1 and 3 will be maintained during normal operations by employing normally open, fused disconnect switches at both ends of the circuit, and the fuses will normally be stored out of the ,
circuit. !
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. . ._ . ._. __m_m_ . . _ . _ _ . . _ . _ . . . _ _ _ _ _ _ _ . . . - . _ . . _ . . _ . . _ _ . . - . . . - . _ .. _ . _ .
I Attichmint 4 PY-CEl/NRR 2352L Pags 62 of 74 Testing of Division 3 equipment is in accordance with the j applicable design bases Table 8.1-2 and in particular Regulatory Guide'1.68. It is designed to permit inspection and testing of all important areas and features, especially those whose operation is not normally demonstrated. As detailed in the Technical Specifications, periodic component tests are supplemented by extensive functional tests during refueling outages, the latter based on actual accident simulated conditions. These tests demonstrate the operability of diesel generator, station battery system components and logic systems and thereby verify the continuity of the system and the operation of components. (Also j refer to Chapter 14 for applicable preoperational tests.)
8.3.1.1.3 Standby Power Sources 8.3.1.1.3.1 Description l
Each division is provided with a diesel engine driven, 4.16 kV, 3 phase, 60 Hz synchronous generator.(see Figure 8.3-1). The diesel generator sets are electrically and physically isolated from each other and are located in a Seismic Category I structure adjacent to the control complex. Note, Division 3 is capable of being manually cross-tied to Division 2 during a station blackout to provide power to some Division 2 loads. Refer to Appendix 15H for more information.4 Figure 8.3-3 shows '
the locations of the standby power sources.
Table 8.3-1 lists loads required for various maximum loading conditions, l
such as loss of offsite power (forced shutdown) and LOCA. The basis for l- the power required for each safety-related load is the motor nameplate rating. In each case this rating is greater than the horsepower required by the driven equipment.
l' 4
- Revision 7 8.3-19 March, 1995 1
f Attichmint 4 PY-CEl/NRR 2352L Page 63 of 74 t
insert - Page 8.3-19 Also note that Division 3 is capable of being manually connected to Division 1 following a loss of coolant accident and a totalloss of both the normal and emergency Division 1 AC electrical power sources, to provide power to the motor operated gate valves in the Feedwater lines.
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Attachmsnt 4 PY CEl/NRR-2352L Page 64 of 74 8.3.1.1.3.3 Division 3 Diesel Generator, High Pressure Core Spray Power Supply
- a. Design Bases The HPCS power system loads consist of the HPCS pump motor and associated 460-volt ac auxiliaries, such as motor operated valves, engine cooling water pump and miscellaneous engine auxiliary loads.
Figure 8.3-7 is the basic one line diagram of the system.
Figure 8.3-8 illustrates the system logic. Table 8.3-9 details the diesel generator specifications.
The HPCS power system is self-contained, except for access to the preferred source of offsite power through the onsite ac power distribution system and the system actuation signal source. The system is operable as an isolated system independent of electrical connection to any other system by using the HPCS diesel generator.
Note, Division 3 is capable of being manually cross-tied to c e.r t Division 2 to provide power to seebe+ tine Division 2 loads during a station blackout. Refer to Appendix 15H for more information.
Class lE auxiliary equipment, such as standby heaters and battery chargers, are supplied from the same source as the HPCS pump motor. l The diesel generator is compatible with power available from the onsite ac power system.
_L.satt The HPCS diesel generator is capable of quickly restoring power to the HPCS pump motor in the event that offsite power is unavailable and can provide all power for startup and operation of the HPCS system. The HPCS diesel generator starts automatically upon receipt of a signal from the plant protection system (low water level or high drywell pressure-LOCA initiation signals) or upon detection of HPCS bus undervoltage or degraded voltage. When the preferred power supply is unavailable, the HPCS diesel generator is automatically connected to the HPCS bus.
Revision 7 8.3-39 March, 1995
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- Attichment 4 PY-CEl/NRR-2352L Page 65 of 74 Insert . Page 8.3-39 Also note that Division 3 is capable of being manually connected to Division 1 following a
' loss of coolant accident and a total loss of both the normal and emergency Division 1 AC electrical power sources, to provide power to the motor operated gate valves in the Feedwater lines.
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Anachm:nt 4 PY-CEl/NRR-2352L Page 70 of 74 pool cooling mode and check that the emergency service water system has been' automatically initiated. After the RHR system and other auxiliary systems are in proper operation, the operator should monitor the hydrogen concentration in the dryvell for proper activation of the I recombiner and mixer, if neces & o pa.%.~ ShoM lai5s de- h hu 4. # h r 6 k g.
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15.6.5.2.2 Systems operations Accidents that could result in the release of radioactive fission products directly into the containment are the results of postulated reactor coolant pressure boundary pipe breaks. Possibilities for all pipe breaks sizes and locations are examined in Sections 6.2 and 6.3, including the severance of small process system lines, the main steam lines upstream of the flow restrictors and the recirculation loop pipelines.
The most severe nuclear system effects and the greatest release of radioactive material to the containment result from a ,
complete circumferential break of one of the two recirculation loop pipelines.
The minimum required functions of reactor and plant protection systems are discussed in Sections 6.2, 6.3, 7.3, 7.6, 8.3, and Appendix 15A.
15.6.5.2.3 The Effect of Single Failures and Operator Errors Single failures and operator errors have been considered in the analysis of the entire spectrum of primary system breaks. The consequences of a LOCA with consideration of single failures are shown to be fully accommodated without the loss of any required safety function. See Appendix 15A for further details.
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Q: g, pg frv q s* cLt k (w r s &M #~) # 5 15.6.6.3.2 Qualitative Results Attachment 4 PY-CEl/NRR-2352L Page 71 of 7<4 The feedvator line break outside the containment is less limiting than either of the steam line breaks outside the containment (analysis
,jer presented in Sections 6.3 and 15.6.4) or the feedvater line break inside the containment (analysis presented in Sections 6.3.3). It is far less limiting than the design basis accident (the recirculation line break analysis presented in Sections 6.3.3 and 15.6.5). !
LNe reactor vessel is isolated on low-low-lov vater level (L1), and HPCS restores reactor water level to the normal elevation. The fuel is covered throughout the transient and there are no pressure or temperature transients sufficient to cause fuel damage.
15.6.6.3.3 Consideration of Uncertainties This event was conservatively analyzed and uncertainties were adequately considered (see Section 6.3 for details).
15.6.6.4 Barrier Performance l
l Accidents that result in the release of radioactive materials outside the containment are the results of postulated breaches in the reactor coolant pressure boundary or the steam power conversion system boundary.
A break spectrum analysis for the complete range of reactor conditions indicates that the limiting fault event for breaks outside the containment is a complete severance of one of the main steam lines as described in Section 15.6.4. The feedvater system pipe break is less severe than the main steam line break.
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15.6.6.5 Radiological Consequences Attachment 4 i
- PY-CEl/NRR-2352L I Page 72 of 74 I 15.6.6.5.1 Design Basis Analysis L
The NRC provides no specific regulatory guidelines for the evaluation of this accident, therefore, no design basis analysis will be presented.
!- 15.6.6.5.2- Realistic Anal'ysis The realistic analysis is based on a realistic, but still conservative !
l assessment of this accident. The specific models, assumptions and the 4
{
program used for computer evaluation are described in Reference 6. 1 t
Parameters used in the evaluation are presented in Table 15.6-20. A schematic diagram of the leakage path for this accident is shown in i Figure 15.6-3.
i 15.6.6.5:2 1 Fission Product Release There is no fuel damage as a consequence of this accident. In addition, an insignificant quantity of activity (compared to that existing in the main condenser hotvell prior to occurrence of the break) is released from the contained piping system prior to isolation closure.
The iodine concentration in the main condenser hotvell is consistent I
with an offgas release rate of 100,000 pCi/see at 30 minutes delay, and is 0.02 (2 percent carryover) times the concentration in the reactor l coolant. Noble gas activity in the condensate is negligible since the
- j. air ejectors remove practically all noble gas from the condenser.
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- 15.6-31
Attichm:nt 4 PY-CEl/NRR-2352L ;
Paga 73 of 74 l 15.6.6.5.2.2 Fission Product Transport to the Environment l The transport pathway consists of liquid release from the break, l carryover to the turbine building atmosphere due to flashing and unfiltered release to the environment through the turbine building ventilation system. I i .
The total integrated mass of coolant leaving the break is 1.454 E6 lbs of condensate. For the purposes of this evaluation, the conservative assumption is made that the activity of iodine per pound of steam is equal to 2 percent of the activity per pound of water.
Taking no credit for holdup, decay or plateout during transport through 1
L the turbine building, the release of activity to the environment is presented in' Table 15.6-21. The total release is assumed to take place l vithin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the occurrence of the break.
l 1
l 15.6.6.5.2.3 Results ;
1
. The calculated exposures for the realistic analysis are presented in Table 15.6-22 and are a small fraction of 10 CFR 100 guidelines.
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15.6.7 REFERCTCES FOR 3E~CTI"6 C fits.6
- 1. Hoody, F. J., " Maximum Two-Phase Vessel Blowdown From Pipes," ASME Paper Number 65-VA/HT-1, March 15, 1965.
- 2. USNRC Standard Reviev Plan, NUREG-75/087.
- 3. Brutschy, F. J., G. R. Hills, N. R. Horton, A. J. Levine, " Behavior of Iodine in Reactor Vater During Plant Shutdown and Startup,"
August 1972, (NED0-10585).
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l Attachment 4 PY-CEl/NRR-2352 Page 74 of 74 Insert Page 15.6-32 15.6.6.5.2.4 Sensitivity Analysis As described in USAR Section 6.2.4.2.2.1.a.1, should a break occur in a feedwater line, the control closure check valves prevent significant loss of reactor coolant inventory and provide immediate isolation. A sensitivity analysis was performed to estimate the amount of leakage l that would have to occur through the control closure check valves in order for the consequences of a feedwater line break outside containment event to exceed the l consequences of the main steam line break outside containment. The results of the l sensitivity ana!ysis are that the leakage through the control closure check valves would have to exceed 200 gallons per minute for each feedwater line (400 gallons per minute total) for i 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order for the consequences of the feedwater line break outside conts:nment to exceed the consequences of the main steam line break outside containment (USAR Table l 15.6-11). The alternative testing performed on these check valves per an exemption to i 10 CFR 50 Appendix J (a visualinspection of the valve seats from undemeath the seat / disc) will verify proper closure of these valves to prevent significant leakage of this order of magnitude.
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Att: chm:nt 5 PY-CEl/NRR-2352L Paga 1 of 2 OPERATOR ACTION EVALUATION
, FOR DESIGN BASIS LOCA AND LOSS OF DIVISION 1 POWER Description Of The Alternate Power Supply Design l
l In order to provide greater assurance that the Feedwater motor-operated valves (MOVs),
B21-F065A and F065B, will be available for closure during a LOCA, a method will be provided to use Division 3 AC power to operate Feedwater MOVs B21-F065A and F065B (normally powered by Division 1). This capability will be provided for the unlikely event of a total loss of Division 1 normal and emergency AC power sources during a LOCA, i.e., Division 1 LOOP and loss of Division 1 diesel power. This alternate power supply is provided by the installation of a power cable between Division 3 (Motor Control Center (MCC) EF1E-1, Compartment V) and Division 1 (MCC EF1 A07, Compartment XT). The cable will be terminated at the load side of a fusible disconnect switch in the MCC compartments at each end. These disconnects will l remain open and the fuses will remain out of their holders until such time as the altemate power supply is needed. The fuses will be stored in the bottom of the MCC compartments so as to be accessible when needed. Labels will also be applied to each compartment's door describing the purpose of the compartment and directing that the fuses are not to be removed from the compartment.
Prior to installing the fuses and closing the breakers for the alternate power supply, operators will be required to open MCC EF1A07's feed breaker at BUS EF-1-A, and all the breakers on MCC EF1 A07. This will prevent potential back feed of power to other circuits. Plant procedures / instructions will direct how to implement this alternate power supply. Only the Feedwater MOVs B21-F065A and F0658 will be operated using this alternate power supply.
The cable and its conduit will be classified safety-related and will be installed with consideration given to seismic and separation concems. The conduit penetrates a rated fire barrier wall between the Division 1 and Division 3 MCC rooms. The penetration area will be restored in accordance with approved procedures.
It has been concluded that the additionalload change for Division 3 under a LOCA condition is within the capabilities of Division 3. It was also determined that the selected cable and fuse sizes are acceptable and that the degraded voltage at the MOVs is within allowable limits.
It is acknowledged that complete separation of Division 1 and Division 3 will not be maintained during use of the alternate power supply. However, supply of Division 3 power will be limited only to the circuits for MOVs B21-F065A/B and then only for a short period of time. The total time that each MOV circuit will actually be energized and not separated is only approximately 68 seconds. Given the limited conditions (only used under emergency circumstances) and time duration for use of the alternate power supply, it is determined that the installation is acceptable from a physical separation and independence aspect.
Once the valves are closed, the Division 1 and Division 3 MCCs will be returned to their original configuration (by procedure) such that if Division 1 power becomes available, the other Division 1 loads supplied from that MCC can be powered.
Summary Of The Operator Action Evaluation This design change results in a potential post-LOCA operator action. It should be noted that I this action is a contingency action in the event that Division 1 power is lost. Therefore, it is not j judged to be a NUREG-0737 action that would need to be taken following all LOCAs. Even so.
this action was evaluated against the considerations in ANSI /ANS 58.8-1984 " Time Response
Attichm:nt S PY-CEl/NRR-2352L Page 2 of 2 Design Criteria For Nuclear Safety Related Operator Actions", to verify that the proposed contingency action can indeed be accomplished.
Severalitems were considered in determining the feasibility of the performance of this operator action. This first was to perform a time-motion study to determine the time required to perform the action in the event of a loss of power. This walkdown showed that the time required to perform all of the steps necessary to start the FWLCS without Division 1 power is approximately 19.5 minutes. This includes the time for the operator to travel from the 599' elevation of the Control Complex, obtain required tools, and perform the required actions, plus time for the Control Room actions to take place. This study showed that the areas required to be accessed to perform this action are readily accessible and can be performed in the time required.
The time considered for the dose evaluation conservatively used the time to perform the entire evolution, even though some of the actions take place in the control room The areas accessed are on the 599 and 620 foot elevations of the Control Complex. All areas requiring access are outside the Radiologically Restricted Area (RRA). Therefore, there are no components (piping, equipment, or other sources)in the travel path or areas of concern that would contain radioactive materials as a result of a Design Basis LOCA that would result in radiation levels that would preclude access to the areas required to perform the proposed action.
Based on a review of calculations and USAR Figures 12.6-2 & 3, the areas to be accessed have average maximum whole body dose rates of s15 mrem /hr. Therefore, the whole body dose to perform this function is approximately 5 mrem. There is the potential for airbome contamination as a result of the LOCA if there is a release in progress. However, the Control Room would be well aware of this and respiratory protection would need to be used as appropriate and in accordance with the guidance of the emergency plan and procedures. The potential addition of the use of respiratory protection has been reviewed with respect to the task being performed. As a result of the simplicity of the task being performed, the use of a respirator is not expected to significantly impact the time required to perform this action.
Therefore, the dose to perform this action is well within the dose limits of s5 Rem or equivalent as required for item II.B.2 of NUREG-0737.
The areas where the actions are being performed all have adequate normallighting and also have battery backed emergency lighting. Therefore, there is adequate lighting in the areas of concem. The environmental conditions in these areas are suitable for a ' mild environment' from an Environmental Qualification (EO) perspective and will consequently result in an environment that is suitable for the operator to perform the required actions. Per the EQ zone drawings which are typically very conservative with respect to personnel access, the maximum temperature in the areas of concern post-LOCA is 82*F. Therefore, there should be nothing to prevent or preclude the operator actions from being performed.
Currently, there is only one action that is required to be performed in the plant in the first hour post-LOCA. This action is to line up backup safety-related air to the outboard MSIV after closure to ensure they remain tightly seated after tf 3 initial accumulator air charge is depleted.
Therefore, with the minimum of two Perry Plant Operators (PPOs) as required by Technical Specification 5.2.2.a, there are adequate personnel to perform the required Feedwater gate valve actions within the time required and within the dose limits of s5 Rem or equivalent as specified by NUREG-0737 and GDC 19.
This operator action outside the control room is assumed to begin 30 minutes after the start of the design basis LOCA, per the guidance of ANSl/ANS 58.8-1984. It then is estimated to take approximately 19.5 minutes to complete. The fill time of the gate valve bonnets is s 9 mmutes.
Therefore, the action of initiating FWLCS can be completed within the currently licensed period of"approximately one hour"(< 64 minutes) following the occurrence of a design basis LOCA, even if normal and emergency Division 1 AC power is not available.
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Attachm nt6 PY-CEl/NRR-2352L Page 1 of 1 SENSITIVITY STUDY RESULTS - FEEDWATER CHECK VALVE LEAKAGE DURING A FEEDWATER LINE BREAK OUTSIDE OF CONTAINMENT USAR Section 15.6.6 addresses the "Feedwater Line Break - Outside Containment" radiologica.
consequence analysis. Sections 15.6.6.3.2 and 15.6.6.4 note that the Main Steam Line Break Outside the Containment analyzed in Section 15.6.4 is the bounding event for breaks outside of l the containment, and that the Feedwater break is less severe than the Main Steam Line Break.
A sensitivity study was performed to determine how much leakage could occur past the Feedwater check valves during a Feedwater line break outside containment, while still remaining ,
bounded by the Main Steam Line Break Outside Containment event. The results of this i sensitivity study will be incorporated into a new paragraph (new USAR Section 15.6.6.5.2.3).
The new paragraph will state:
"As described in USAR Section 6.2.4.2.2.1.a.1, should a break occur in a feedwater line, the !
i control closure check valves prevent significant loss of reactor coolant inventory and provide immediate isolation. A sensitivity analysis was performed to estimate the amount of leakage that would have to occur through the control closure check valves in order for the consequences of a ,
feedwater line break outside containment event to exceed the consequences of the main steam ;
lirie break outside containment. The results of the sensitivity analysis are that the leakage through the control closure check valves would have to exceed 200 gallors per minute for each feedwater line (400 gallons per minute total) for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order for the consequences of the feedwater line break outside containment to exceed the consequences of the main steam line break outside containment (USAR Table 15.6-11). The attemative testing performed on these ;
check valves per an exemption to 10 CFR 50 Appendix J (a visualinspection of the valve seats '
from undemeath the seat / disc) will verify proper closure of these valves to prevent significant leakage of this order of magnitude."
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Attichm:nt 7 PY-CEl/NRR-2352L Pags 1 of 3 RISK INFORMED REVIEW OF THE FEEDWATER CONFIGURATION The proposed addition of both divisions of FWLCS to the bonnets of both of the gate valves on the Feedwater lines, together with the provisions for providing alternate power from Division 3 over to the gate valves, has been evaluated and shown to improve the reliability of the Feedwater penetrations. A risk-informed Probabilistic Safety Assessment (PSA) approach was utilized to examine, on a relative basis, the current versus the proposed containment isolation provisions for the Feedwater lines.
This assessment concentrates on the function of the FWLCS to prevent fission products bypassing containmmt following a design basis loss of coolant accident (LOCA). However, due to the nature of the PSA process, it also examined events beyond the design basis LOCA. A Level 1 core damage frequency (CDF) analysis is normally utilized to determine if a proposed change is acceptable. In this case, the CDF would not be changed at all (certainly not in an adverse manner) by the proposed revision to the isolation provisions on Feedwater. Therefore, this analysis was primarily a Level 2 containment isolation type analysis, since the Level 1 CDF analysis results (no increase in core damage frequency) would clearly permit implementation of the proposed changes.
In addition to the core damage frequency being unaffected by this change, a Level 1 CDF discussion is relevant in other ways in determining the overall acceptability of the proposed configuration. Thi* 4 because the Feedwater penetration does not need to be sealed by the Feedwater Ls.akage Control System unless core damage has occurred. An accident that proceeds to the point of core damage is a very low probability 4
event. For PNPP, the core damage frequency from a large break LOCA is 3.05x10 per year, a very low val se. The combined core damage frequency from all 4
the LOCAs (spectrum from small to large) 's approximately 5x10 per year, also very low. It shou,'d also be noted that if the LOCA is in any line other than the feedwater line, the feedwater lines will likely remain filled with feed water, preventing radiological leakage, and upon vessel reflood, the lines will refill with water, preventing air leakage.
This water seal would even exist for many of the postulated break locations in the Feedwater lines, i.e., those that are not at the low point in the piping inside the containment. For such cases, the Feedwater Leakage Control System is superfluous (i.e., not necessary) due to the water seal on the penetration provided by the initial water or the reflood water. The com damage frequency from such a feedwater line 4
break at the low point of the system would be a small subset of the 5x10 per year combined CDF for the spectrum of LOCAs described above.
Most any other events that could lead to core damage also would have a pressurized vessel at the time that Feedwater stops feeding the vessel (events such as anticipated transients without scram, various other transients, luss of offsite power, station blackout, etc.). This pressure would provide a strong seating force on the check valves in the line at the beginning of the event, which is how the valves are designed to seal well. Many of these events would also provide the water seal on the Feedwater penetration since the Feedwater line would not be broken, and water would remain from the initial injection feedwater or the reflood water. Regardless, the frequency of these core damage events that might require isolation of the feedwater lines is also very low:
Attachment 7 PY-CEl/NRR-2352L Page 2 of 3 i
e ATWS 2.26x10 4 l
- Total for various other transients 3.80x10 4
- LOOP 5.68x10 4
- Station Blackout 2.28x10 4 Although PNPP has not performed a Level 2 containment PSA, this proposal was examined to determine whether the reliab;lity of the Feedwater line isolation would be improved or reduced. The assessment clearly showed that the proposed configuration will provide a i significantly higher probability that if needed (i.e., the frequency of a significant core damage I
event is treated as unity (1)), the 'FWLCS will effectively seal the line within the one hour time frame in the existing licensing basis. It also shows that the probability is increased that the l high integrity, low leakage vaives that are credited in the offsite dose calculations (the third (motor-operated gate) valves) will indeed be closed.
The establishment of an effective water seal on the penetration from the FWLCS within an hour for the current configuration is contingent on prompt operator actions. This is because with the current installed system, it takes from 36 to 44 minutes for the system to fill the entire piping volume to a level covering the seats of the valves. This need to initiate the system promptly (assumed to be at 20 minutes after the event) to allow for this fill time dominates the reliability analysis. Even if credit is given in the evaluation of the current configuration that the check valves will successfully permit the filling of the pipe, the reliability l analysis shows a tremendous improvement in reliability for the new proposed design, due to l the extra time that is available to the operator before the system must be initiated (the new '
system establishes an effective seal in s 9 minutes, so the operator has up to 55 minutes to recognize the need for and initiate the system).
Also, the new design includes a pre-existing alternate power supply from Division 3 over to l Division 1 to power up the third (motor-operated gate) valves on each Feedwater line in the event of a totalloss of both the normal and emergency AC power source to these valves.
This results in an improvement in the reliability of closure of thu high integrity valves.
Fault tree analyses were performed for the existing configuration and the proposed modification. Each of these fault trees was quantified under two sets of boundary conditions to reflect the likelihood of system failure under design basis (LOCA with a Luss of Offsite Power (LOOP)) constraints as well as a LOCA without a LOOP.
The Fault Tree Analysis of the system used the WinNUPRA code. Component reliability data was obtained from the PNPP Individual Plant Examination (IPE) model. Human reliability was modeled using the Electric Power Research Institute (EPRI) Systematic Human Action Reliability Procedure (SHARP).
The following assumptions were made in performing the analysis (results are provided in l Table 1):
l
- 1. For the current configuration (the baseline analysia, with credit given for the inboard FWLCS injecting between the two check valves), the check valves are considered capable (subject to random failure to close) of acting as " poppets" in order to maintain a i
l l Attichmint 7 I
PY-CEl/NRR-2352L l l P gs 3 of 3 l water inventory and positive seal in the interspace between the isolation valves.
l 2. Regarding the FWLC system injection, the start and run failures of the pumps are l assumed to be the dominant contributors to failure to provide water to the space between i the isolation valves. Other pump or valve failures are not modeled for the water injection pathway.
- 3. The existing F065A and B valves derive power from Division 1. For the " proposed design" case, an alternate power supply for the F065 A and B valves is also assumed, l from Division 3.
- 4. The probability of the operator failing to initiate closure of the isolation MOVs and initiate FWLCS (known as the Human Error Probability (HEP)) is 0.26, given action is required in the first 20 minutes of the accident (baseline case). The probability of the operator failing to perform the action within 55 minutes (" proposed design" case) is a HEP of 0.036. For the fault tree branches that involve the need to use the alternate power supply, the operator action to align the power supply to the B21-F065 valves ;s a HEP of 0.067.
TABLE 1 The probability of failure to establish a water seal within one hour of a Core Damage Accident is:
Offsite Power Available With LOOP Current Design
- 0.267 0.28 Proposed Design 0.0419 0.0689
- for the " Current Design" numbers, it was conservatively assumed that a water seal between the check valves, with the gate valve open, is a success".
When the " Current Design" case was instead examined assuming that a success was
- dependent on closure of the third (motor-operated gate) valves, the failure probability was higher than the Current Design numbers presented in Table 1 above. The " Proposed Design" case presented above already assumes that closure of the third (gate) valves is necessary for success.
As noted above, the conclusion of the assessment is that in general, the human operator error probability on initiation of FWLCS dominates the results. The primary reduction in system unreliability is seen from the reduction in the FWLCS fill time from 40 minutes to a l few minutes. This is observed in the order of magnitude improvement in system reliability l between the baseline case (current configuration) and the proposed alternate system l configuration.
Given the substantial reduction in system failure probability when the alternate Case is evaluated, it was determined appropriate to pursue the FWLCS modification to utilize the isolation MOV as the water seal, and to allow for a redundant power supply to the MOV.
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Attechment 8 1
PY-CEl/NRR-2352L
'? age 1 of G RE_ QUEST FOR EXEMPTION
SUMMARY
Pursuant to the requirements of 10 CFR 50.12, an Exemption is requested to 10 CFR 50 Appendix J for the Perry Nuclear Power Plant (PNPP), regarding the testing methodology for the Feedwater penetration check valves. The Exemption request would permit use of an alternative testing methodology per the Inservice Testing Program (Technical Specification 5.5.6), consisting of a visualinspection of the valve internals from under the seat to verify proper closure. This alternative testing methodology would ensure that the Feedwater check valves are continuing to close properly to provide a containment isolation function.
This proposed exemption request supports the use of this alternate testing methodology. The proposed change is based on a combination of factors incorporating consideration of the design basis, licensing basis, deterministic evaluations, risk-informed evaluations, cost-benefit analyses, and design changes that are proposed for implementation on the Feedwater penetrationc to improve their overall reliability.
Further details on these considerations is provided below in the exemption request. In summary, the information provided notes that:
The Feedwater check valves will continue to be treated as safety-related containment isolation valves,
. They will be tested per the Inservice Testing Program, Proper operation and closure of the check valves will be assured, When disassembled in the past, these check valves have not shown damage to the seats that would require repair, The original manufacturers tolcance specifications have been verified on these valves, A possible cause of the testing difficulties may simply be operational conditions preceding the test date, i.e., low flows from the RWCU or RHR systems through the valves in the open direction, so the valves are not stroking closed from a full open position, combined with the i Iow testing pressures and low volumetric flow that occurs during testing at PNPP, ,
Post-accident dose calculation assumptions continue to be met by the excellent as-found leak I test results achieved by the motor-operated gate valves installed in the Feedwater penetration for the purpose of providing long-term, high integrity leakage protection, as provided in the licensing basis for PNPP, A water seal on the Feedwater penetrations will still be provided following a postulated core damage event by one of several methods: the Feedwater piping inside containment for any core damage accident other than a break in the Feedwater line at the lowest part of the piping in the drywell, the water seal provided by the relocated Feedwater Leakage Control System on ,
the seat, stem and bonnet of the Feedwater gate valves, or the water in the Feedwater system l piping outboard of the Feedwater gate valves that willlikely remain intact, The design changes being proposed for the Feedwater penetrations result in an improvement l in the ability to isolate the line within the "approximately one bour" time frame that is in the current licensing basis, as demonstrated by a reliability analysis incorporating human error probability considerations, the relocation of the FWLCS to the bonnet of the gate valves, and the availability of an alternate power supply, and
. Continued leak rate testing and subsequent rework would result in expenditure of additional dose and dollars that are not justified on a cost-benefit basis.
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Attschment 8 PY CEl/NRR-2352L Page 2 of 6 CRITERIA FOR GRANTING AN EXEMPTION REQUEST in accordance with 10 CFR 50.12, " Specific Exemptions" the Commission is authorized to grant an Exemption upon a demonstration that the Exemption: (a) is authorized by law, (b) will not present an undue risk to the public health and safety, and (c) is consistent with the common defense and security, in addition, the Commission will not consider granting an Exemption unless one of the following special circumstances exist: (i) application of the regulation in the particular circumstcnces conflicts with other rules or requirements of the Commission, (ii) application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule, (iii) compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated, (iv) the Exemption would result in benefit to the public health and safety that compensates for any decrease in safety that may result from the granting of the Exemption, (v) the Exemption would provide orily temporary relief from the applicable regulation and the licensee has made good faith efforts to comply with the regulation, or (vi) there is present any other material circumstance not considered when the regulation was adopted for which it would be in the public interest to grant an Exemption.
The Exemption is Authorized by Law The Commission's authority to grant Exemptions from its regulations is defined in Title 10 of the Code of Federal Regulations. Specifically, the requirements of 10 CFR 50.12 are applied to this specific Exemption request. Therefore, this Exemption request is consistent with the regulatory scheme established by the NRC and is not prohibited by any statutory authority. Hence, the Exemption may be authorized under NRC regulations. Therefore, the Exemption is authorized by law.
The Exemption Will Not Present an Undue Risk to the Public Health and Safety An Exemption request will not present an undue risk to the public health and safety if it can be shown that the Exemption meets the statutory standard of adequate protection to the health and safety of the public.
The evaluation of"no undue risk" considers such factors as the type of plant operation contemplated, the existence of alternative methods of compliance or compensatory measures which meet the objective of the regulation (or requirement) but in a different way from that originally envisioned, and other safety factors.
The type of plant operation contemplated is simply the performance of an alternate test method for verifying propor operation and closure of the Feedwater system containment isolation check valves. The alternative method of compliance or compensatory measure which meets the objective of the regulation but in a different way from that originally envisioned consists of the visualinspection of the seats of the Feedwater check valves in accordance with the inservice Testing Program (PNPP Technical Specification 5.5.6). The valves will still be considered
Attachment 8 PY-CEIMRR-2352L Page 3 of 6 safety-related containment isolation valves. It will still be necessary to ensure that the check valves are properly closing. As noted in Generic Letter (GL) 89-04 Position 3 "Back Flow Testing of Check Valves", many plants were not performing back flow testing of any kind prior to 1989 on check valves that perform a safety function in the closed position to prevent reversed flow. Main Feedwater check valves were listed as an example of such valves. The GL noted that tests on check valves that perform a safety function in the closed position to prevent reversed flow should be performed. " Category C" teste on such " safety function check valves" were described as acceptable if they prove that the disc closes on its seat. Two different example methods discussed for such safety function Category C tests were leak rate testing or a visual observation. Upon implementation of this amendment, the PNPP Feedwater check valve closure function will be verified by an inspection of the valves, from a location under the seat, per the Inservice Testing Program (Technical Specification 5.5.6). Design changes will be made to incorporate new taps off of the line adjacent to the check valves so that this necessary inspection can be completed (for example, by using boroscopic inspections). The chek valves would be removed from the hydrostatic testing program, as cuch additional testing is not necessary to ensure their closure capabilities to prevent significant reverse flow.
These check valves have taen disassembled during several refueling outages for inspection and maintenance due to difficulties experienced during the leak rate testing that is currently required on the valves. The check valve seats have not been damaged or shown signs of needing to be reworked. There was no evidence that plant operation was degrading the metal faced seats on these piston style check valves. However, rather than simply retesting the valves, rework (polishing or relapping of the seats) has been performed during the past several refueling outages as a corrective action for the test failures. When the check valves need to be disassembled and polished or relapped, it is a major task. The valves are in a very tightly confined area, with limited access for valve disassembly and rework. The lapping / polishing equipment must be partially disassembled for installation into the inboard [1N27F559A/B]
valves. The dose rates in the area of the valves inside the Drywell are high in comparison to most work areas. During the most recent refueling outage (RFO6), over 5 Rem (5000 mrem) was received by workers doing feedwater check valve seat work. Due to the valve's installed configuration, the worker doing inspections after relapping must crawl down into the valve from above and perform the inspections while he/she is upside down and in a contorted position, in order to get out, helpers must pull the worker up out of the valve.
By RFO5, it was decided to utilize Edward Valves incorporated for the valve work to gain access to the manufacturer's design drawings and other proprietary information. Therefore, Edward Valves performed the RFOS machining using special tools designed specifically for these valves. After machining, the discs were blue checked under the direction of the Edward Valves representatives by rotating the disc 120 degrees three times. This was done to address the potential for disc rotation. Despite these efforts, the valves did not perform significantly different during leak rate testing in RFO6. Edward Valves employees were again involved in the RFO6 rework of the valves.
Although the seats have been restored to the manufacturers original dimensional specifications, and the above described disassemblies and inspections have shown that the valves do not show evidence of seat damage from plant operation, as-found testing difficulties continue. A possible cause of the testing difficulties is simply the plant operational conditions that precede the test date, i.e., the running of low amounts of water flow from the RWCU or RHR system through the valves in the open direction during the period following plant shutdown, so that the valves do not stroke closed from a full open position, where they would benefit from the gravity
l Attachment 8 l
PY-CEl/NRR-2352L Page 4 of 6 l
assist. This is not representative of post-accident conditions. Also, the subsequent testing is performed at very low differential pressures, due to the low peak post-accident pressure (Pa) at a BWR-6 plant like PNPP. The testing is also performed using a very low volumetric flow rate l (the input volume is the amount of flow through a %" line). These low pressures and low flows do not provide a strong re-seating force on the valves. Again, there has not been evidence that these valves actually benefit from the dose and money spent in performing rework of the seats.
Post-accident dose calculation assumptions continue to be met by the excellent as-found leak test results achieved by the motor-operated gate valves installed in the Feedwater penetration for the purpose of providing long-term, high integrity leakage protection, as provided in the licensing basis for PNPP. This licensing basis, which shows that the check valve leakage is not considered to contribute to the offsite dose calculations, is summarized in Attachment 1 to this letter, which notes the letters, the USAR Sections, and the NRC approval documents that reflect this current licensing basis. The alternate test proposed by this amendment will provide confidence that these Feedwater check valves are operating and closing properly on an ongoing basis.
A risk-informed discussion of the proposed license amendment and exemption request is provided in Attachment 7 to this letter. Five aspects of the Attachment 7 discussion that are relevant to this exemption request are that:
- 1. the Core Damage Frequency (CDF) would not be changed at all(certainly not in an adverse manner) by the proposed revision to the isolation provisions on Feedwater
- 2. the Feedwater penetration does not need a water seal unless a core damage accident has occurred that requires containment isolation (such an accident is a very !ow probability event)
- 3. there is a high likelihood that a water seal will be maintained on the Feedwater lines following a postulated core damage event, even if the FWLCS is not providing that seal
- 4. many events that could lead to core damage also would have a pressurized vessel at the time that Feedwater stops feeding the vessel, which would provide a strong seating force on the check valves in the line at the beginning of the event, which is how the valves are designed to seal well, and,
- 5. When the penetration isolation system reliability was examined using a risk-informed process, an order of magnitude improvement in system reliability was seen between the baseline case (current configuration) and the proposed alternate system configuration.
Details on the core damage frequencies (CDFs) are in Attachment 7 for the first two items above.
1 For the third item, Attachment 7 notec that if the event in question is not a break of the l Feedwater lines, there will likely remain a water seal on the penetration. If the LOCA is in any line other than the Feedwater line, the Feedwater lines will likely remain filled with feed water, preventing radiological leakage, and upon vessel reflood, the lines will refill with water, preventing air leakage. This water seat would even exist for many of the postulated break locations in the Feedwater lines, i.e., those that are not at the low point in the piping inside the i containment. For such cases, the Feed. water Leakage Control System is superfluous (i.e., not necessary) due to the water seal on the penetration provided by the initial water or the reflood water. The core damage frequency from such a line break at the low poird of the Feedwater j
system would be a small subset of the 5x10-8 per year combined CDF for the spectrum of LOCAs postulated at PNPP. Although it is not discussed in Attachment 7, the licensing basis i for PNPP also makes a strong case that the feedwater piping outboard of the motor-operated
Attachm:nt 8 PY-CEl/NRR-2352L Page 5 of 6 gate valves would remain intact to provide a water seal on the Feedwater lines. Although the NRC only explicitly credited this piping integrity for the first hour following an event there is a high likelihood that this piping would remain intact for the duration of a core damage event. This
! is piping designed to operate at high pressures and withstand various thermal stresses and I
movements. All stress levels in the ANSI B31.1 portion of the Feedwater piping are below the ANSI B31.1 allowables.
Details on the fourth item are provided in Attachment 7.
i I For the fifth item, the design changes being proposed for the Feedwater penetrations result in an improvement in the ability to isolate the line within the "approximately one hour" time frame that is in the current licensing basis, as demonstrated by a reliability analysis incorporating human error probability considerations, the relocation of the FWLCS to the bonnet of the gate valves, and the availability of an alternate power supply. This reliability evaluation is also summarized in Attachment 7 to this letter, and in the following discussion.
The design changes proposed for the Feedwater penetrations involve relocation of the Feedwater Leakage Control System piping over to the bonnet and seat of the third (gate) valves on each of the two Feedwater lines, and permanent provisions for an alternate power supply from a redundant AC Division to power the gate valves in the event of a totalloss of their normal and emergency AC power supply from Division 1. The assessment of the proposed Feedwater configuration, including the design changes, clearly showed that the proposed configuration will provide a significantly higher probability that if needed (i.e., the frequency of a significant core damage event is treated as unity (1)), the FWLCS will be able to be effectively sealing the line within the one hour time frame in the existing licensing basis. It also shows that the probability is increased that the high integrity, low leakage valves that are credited in the offsite dose calculations (the third (gate) valves) will indeed be closed.
The conclusion of the assessment is that in general, the human operator error probability on initiation of the FWLC system dominates the results. The primary reduction in system unreliability is seen from the reduction in the FWLCS fill time from approximately 40 minutes to a few minutes. This is observed in the order of magnitude improvement in system reliability between the baseline case (current configuration) and the proposed alternato system configuration.
Given the substantial reduction in system failure probability when the alternate case is evaluated, it was determined appropriate to pursue this FWLCS modification to utilize the isolation MOV as the water seal, and to allow for a redundant power supply to the MOV.
Therefore, based on the combination of the above considerations, the granting of the requested Exemption will not present an undue risk to the public health and safety.
The Exemption is Consistent with the Common Defense and Security l
With regard to the " common defense and security" standard, granting the requested Exemption is consistent with the common defense and security of the United States. The Commission's Statement of Considerations in support of the Exemption rule note with approval the explanation of this standard as set forth in Long Island Lighting Company (Shoreham Nuclear Power Station, Unit 1) LBP-84-45,20 NRC 1343,1400 (October 21984). There, the term " common c
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Atttchment 8 PY-CEl/NRR-2352L Page 6 of 6 defense and security" refers principally to the safeguarding of special nuclear material, the absence of foreign control over the applicant, the protection of Restricted Data, and the availability of special nuclear material for defense needs. The granting of the requested Exemption will not affect any of these matters and, thus, the requested Exemption is consistent ;
with the common defense and security.
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Special Circumstances Are Present I Special circumstances exist for the proposed Exemption request as prescribed by 10 CFR 50.12(a)(2)(ii) and (iii), specifically, application of the regulation in the particular circumstances is not necessary to achieve the underlying purpose of the rule, and compliance would result in undue hardship, respectively.
Applying the current regulation (10 CFR 50 Appendix J) in the particular circumstance is not necessary to achieve the underlying purpose of the rule since the valves would continue to be tested using the Generic Letter 89-04 Category C test for Feedwater check valves which perform a safety function in the closed position to prevent reverse flow. As explained in the Generic Letter, the test needs to prove that the disc closes on its seat. An example method discussed for such safety function Category C tests is a visualinspection, such as would be
- implemented on the PNPP Feedwater check valves upon approval of this amendment and exemption request. This inspection would be controlled by the PNPP Inservice Testing Program (Technical Specification 5.5.6). This would ensure that the check valves are working properly, and are properly seating.
l Also, compliance with Appendix J by performing leak rate tests on these valves rather than performing the attemate visual inspection on these Feedwater penetration check valves would result in an undue hardship. As noted in Attachment 1 to this letter, leak testing and the subsequent reworking of these valves resulted in approximately 5 rem of dose to plant workers l in the most recent refueling outage, and approximately $880,000 was spent in performing the maintenance work. As noted above, these dose and monetary expenditures appear to be of little value, since the seats have been restored to the manufacturers original dimensional specifications, the valves do not show evidence of seat damage when they are disassembled, and r nossible cause of the testing difficulties is simply the plant operational conditions that pre %e the test date, i.e., the running of low flow from the RWCU or RHR systems through the valves in the open direction, se that the valves are not stroking closed from a full open position ,
where they would benefit from the gravity assist, combined with the limitations on the testing l pressures and volumetric flow rates.
Continued expenditures on this order (5 rem /$880,000 per outage) are not warranted. This dose can be avoided and the funds can be allocated for more beneficial projects if the proposed j alternate test method of a visual inspection of the valve seats from under the seat area is approved.
CONCLUSION I Because the Exemption is authorized by law, will not present an undue risk to the public health and safety, is consistent with the common defense and security, and special circumstances exist,10 CFR 50.12 authorizes the Commission to grant this specific Exemption.
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