ML18037A039
| ML18037A039 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 09/21/1984 |
| From: | Mangan C NIAGARA MOHAWK POWER CORP. |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| (NMP2L-0165), (NMP2L-165), NUDOCS 8409250352 | |
| Download: ML18037A039 (117) | |
Text
DOCKET 05000410 ACCESSION NBR:8409250352 DOC ~ DATE: 84/09/21 NOTARIZED;,YES FAOIL:50-410 Nine Mile Point Nuclear Stations Unit 2i Niagara Moha AUTH,NAME AUTHOR AF F ILIATION MANGANgC ~ VS Niagara Mohawk Power Corp'ECIP
~ NAME RECIPIENT AFFILIATION SCHNENCERiA, L,icensing Branch 2
R
<'UBJECT:
Forwards resPonses to SER OPenItems 120r107r181r182ep142b r
182cii83r 184 L 185 'esponses will be, included in next FSAR.
amend,N/16 oversize tables'perture cards available'n-POR<
DISTRIBUTION CODE:
8001D COPIES RECEIVED:LTR ENCL
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NIAGARAMOHAWKPOWER CORPORATION/300 ERIE BOULEVARDWEST, SYRACUSE, N.Y. 13202/TELEPHONE (315) 474-1511 September 21, 1984 (NMP2L 0165)
Mr. A. Schwencer, Chief Licensing Branch No.
2 U.S.
Nuclear Regulatory Commission Washington, DC 20555 Re:
Nine Mile Point Unit 2 Docket No. 50-410
Dear Mr. Schwencer:
Enclosed for your use and information are the Nine Mile Point Unit 2 responses to the Nuclear Regulatory Commission's Safety Evaluation Report open items.
This information has been previously discussed with your staff and is submitted to aid your review of the Unit 2 license application for the resolution of these open items.
This submittal includes information for Safety Evaluation Report open items
- 120, 147,
- 181, 182a,
- 182b, 182c,
- 183, 184, 185.
The enclosed wi 11 be included in the next Final Safety Analysis Report Amendment.
Very truly yours, NLR:ja Enclosure xc: Project File (2)
C.
V.
Man n
Vice President Nuclear Engineering
& Licensing 840925O352 84092i voa aoaCK osoooexo
'E'
'DR
'r 1< '4 f
pl
~r III
UNITEO STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Hatter of Niagara Mohawk Power Corporation
)
(Nine Mile Point Unit 2)
)
Oocket No. 50-410 AFFIDAVIT C.
V. Mangan, being duly sworn, states that he is Vice President of Niagara Mohawk Power Corporation; that he is authorized on the part of said Corporation to sign and file with the Nuclear Regulatory Coranission the documents attached hereto; and that all such documents are true and correct to the best of his knowledge, information and belief.
Subscribed and sworn to before me, a Notary Public in and for the State of New York and County of ~
a.
, this g/-
day of m~
1984.
Notary Public in and for ncaa a,
- County, New York My Coranission expires:
CHRISTINE AUSTIN Notarv Pyblic in the State of Net Yorft ualiFed in Onondaga Co. No. 47876&g y Commission Exprres March SO, l~
WfTNA3N15Pi%
AYAN lo 8JS e9 ni o:tdo'I e".!o)l
$53TQCN 4;~A.o9 qstnc,".0 ni k~Ckd)
Ml,QE Crt'eugQ a~mme'3 pe
5Q.~126)
Nine Mile Point Unit 2 FSAR QUESTION F470.7 (12.4)
Section 12.4 indicates that the dose 'ssessment is in progress and that doses and details of the man-rem evaluation will be provided in an amendment.
The bases and details of the dose assessment, as specified in Regulatory Guide 1.70, Revision 3, and Standard Review Plan 12.3-12.4 (NUREG-0800),
should be supplied.
Either provide this information or a schedule for submitting the information.
RESPONSE
See revised Section 12.4.
- 1. 12
- 1. 13
- l. 15 1.17 1.18 1.21 1.22 j I
~8gpg250352 Amendment chl2 177fqr14x QEcR, F470. 7-1 09/05/84 155
P
Nine Mile Point Unit 2 FSAR QUESTION F471.11 Subsection 12.4.2.1, Man-Rem Evaluation, states that the Man-Rem evaluation is in progress, the details of which will be included as an amendment to the Unit 2 FSAR.
Provide this information.
RESPONSE
See revi,sed Section 12.4.
1.10 1.11
- 1. 12
- 1. 14 Amendment, QEcR F471.11-1 chl2177fqr14y 09/05/84 155
Nine Mile Point Unit 2 FSAR 12.4 DOSE ASSESSMENT
- 1. 10 Radiation exposures in the plant are primarily from components and equipment containing radioactive fluids, and to a
lesser extent from the presence of airborne radionuclides.
Inplant radiation exposures during normal operation, refueling, and anticipated operational occurrences are discussed in Section 12.4.2.
Radiation exposures at onsite locations outside the plant are discussed in Section 12.4.3.
1.12 1.13 1.14
- 1. 15
- 1. 17 12.4.1 Design Criteria
- 1. 19 The criteria for doses to plant personnel during normal operation and anticipated operational occurrences, including refueling,.
are based on the requirements discussed in 10CFR20.
The design radiation lev'els during normal operation, refueling, and anticipated occupational occurrences are shown on Figures 1'2,3-34 through 12.3-66.
1.20 1.21 1.22
- 1. 24
- 1. 25
- 1. 26 Radiation exposures to operating personnel are within 10CFR20 limits.
Radiation protection design features (Section 12.3) and the health physics program (Section 12.5) assure that the occupational radiation exposures (ORE) to-operating personnel during normal operation, refueling, and anticipated operational occurrences are as low as is reasonably achievable (A?ARA).
1.27 1.28 1.30 1.31 12.4.2 Exposures Within the Plant 12.4.2.1 Man-Rem Evaluation
- 1. 40
- 1. 42 The occupational radiation dose assessment for.Unit 2 is performed using the guidelines of Regulatory Guide 8.19'i'.
The bases for the annual man-rem estimates are Unit 1
operating data are modified to account for differences and improvements in Unit 2.
The projected radiation dose rates throughout the plant facilities are based on assumed radiation conditions after 5 yr of plant operation and expected radiation dose rates.
Operational data from several BWRs'~'which show that the average annual man-rem per unit over several operating
= years is 948 man-rem per year) are presented in Table 12.4-12.
These data indicate that, in recent years, occupational.
radiation-exposures have. been much larger than the radiation exposures reported for operating BWR plants in the mid-1970's.
The primary reason for the increase in radiation exposure. has been the increase in manpower necessary to support the expanding special maintenance activities.
- l. 45
-1. 46 1.48 1.49 1.50 1.51 1.52 1.53
- 1. 54 1.55 1.56 1.57 Amendment
- 12. 4<<1 ch1217718 f-14cy 09/05/84 155
Nine Nile Point Unit 2 FSAR Table 12.4-13 shows the distribution of annual occupational radiation exposures by work functions for a all BWRs over several years as suggested in Regulatory Guide 8.19.
The average values indicate that operating BWR plants have approximately 76 percent annual occupational exposure attributed to routine maintenance (40 percent) and special maintenance (36 percent).
In recent
- years, plant modifications attributed to feedwater sparger
- repairs, inspection, repair and replacement of recirculation piping, TNI lessons-learned modifications, and increased snubber and pipe hanger inspections have contributed to the growing amount of occupational radiation exposures associated with special maintenance work functions.
Design features described in, Sections
- 12. 1 and 12.3 for the Unit 2 BWR 5 plant should minimize the special maintenance work experienced. at earlier-designed operating BWR plants.
1.58 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 2.10 2.11 Unit 2 design improvements that are expected to reduce the occupational radiation exposures include the following:
2.12 2,13 Incorporation of flush connections, on the CRD scram discharge volume header permits condensate flushing of piping to minimize corrosion product holdup in a high personnel access area.
- 2. 15 2.16 2.17 2.
3.
Use of filtered'ondensate water for CRD hydraulic fluid and the reactor recirculation pump seal purge provides a
clean water source that should extend pump seal life.
Installation of permanent hoisting system and access platforms for the recirculation
- pumps, main steam isolation
- valves, and safety-relief valves minimizes maintenance time in the drywell.
2.18 2.19 2.20 2.21 2.22 An improved refueling platform makes fuel handling activities more efficient, therefore less time is spent on the platform.
A multistud tensioner reduces the amount of man-hours necessary to handle the reactor vessel head studs.-
2.23 2.24 2.25 2.26 6.
A new handling tool and platform for the removal of CRDs from beneath the reactor vessel reduces crew size and time spent in the high radiation area.
2.27 2.28 7.
Improved fuel design minimizes the buildup of radiation levels near reactor coolant systems and 2.29 2.30 Amendment.
- 12. 4-la ch1217718f-14cy 09/05/84 155
Nine Mile Point Unit 2 FSAR reduces the amount of fuel assembly sipping activities.
8.
Improved piping material for the recirculaiton system eliminates the special maintenance that was required on older BNR recirculation piping due to stress corrosion cracking.
2.31 2.32 9.
10.
Inservice inspection access is improved'y remote equipment development and access doors for reactor vessel and nozzle weld inspection.
The main steam isolation valves are ball valves with greatly reduced maintenance requirements and smaller leakage rates which reduce the amount of man-hours spent servicing and inspecting the valves.
2.33 2.34 2.35 2.36 2.37 11.
A decontamination platform is provided to wash the walls of the reactor cavity pit and internals pool to minimize contribution from this source.
2.38 2.39 12.
Use of separate shielded cubicles for locating redundant components and highly radioactive components minimizes radiation exposures during maintenance activities.
2.40 2.41 13.
Use of mechanical snubbers should reduce the frequency of necessary inspection compared to hydraulic-.operated snubbers.
2.42 2.4 14.
Installation of a
CRD flush tank removes highly radioactive corrosion and fission products from CRD internals prior to rebuilding.
2.4 2.4 The occupational radiation exposure for Unit 2 is determined for each of the Regulatory
. Guide 8.19ork function categories by identifying specific tasks within each of the seven work function categories and determining the time and manpower requirements for those tasks.
This information is used with the expected dose rates in areas where work is performed to determine the radiation exposure from each activity.
Tables
- 12. 4-5 through
- 12. 4-11 provide estimates of occupational exposures based-on the identification of specific tasks within each of the seven work function categories:
routine operations and surveillance, nonroutine operations and surveillance, routine maintenance, radwaste processing, refueling, 'nservice inspection, and special maintenance.
Table 12.4-4 summarizes the occupational dose estimates for the seven work functions.
A comparison
- 2. 47 2.48 2.49 2.51 2.52 2.53 2.54 2.55 2.56 2.57 2.58 3.1 Amendment
- 12. 4-1b ch1217718f-14cy 09/05/84 155
Nine Mile Point Unit 2 FSAR between Tables 12.4-4 and 12.4-12 shows that the Unit 2 occupational exposure is consistent with the operating plants data for the period of 1974-1979 (before the TMI accident).
The higher occupational exposures for the period of 1980-1982 are not expected at Unit 2 because plant modifications that caused the increases have been incorporated into the original design of Unit 2.
12.4.2.2 Estimates of Inhalation Thyroid Doses Inhalation doses during full-power operations will be negligible in every.area except the
- reactor, turbine, and radwaste building areas.
Potential airborne activities for these areas are given in Section 12.2.2.
These concentrations are based upon data given in NUREG-0016 and EPRI'-495.'he inhal'ation thyroid doses that.
result are given in Table 12.4-2.
Thyroid dose rates in Table 12.4-2 are calculated according to:
D
= $ (B.R.)
(A' (C)
(K
)
l 3.2 3.3 3.4 3.5 3.6 3.8 3.10
- 3. 13 3.14 3.16 3.17
- 3. 19 Amendment 12.4-lc chl217718f-14cy 09/05/84 155
I )
0
Nine Mile Point Unit 2 FSAR References
- 1. 10 1.
Regulatory Guide 8.19>
Occupational Radiation Dose Assessment in f.ight-Water Reactor Power Plants Design Stage Man-Rem Estimates Revision 1,
June 1979.
1.12 1.13 2.
NUREG-0713, Volume 6>Occup<<tiorr'r I. Radiation Exposure at
- 1. 14 Commercial Nuclear Power Re rctor 1902, December 1903.
Amendment
- 12. 4-4 ch12 177 18f-14dk 09/05/84 155
Nine Mile Point Unit 2 TABLE 12. 4-4 ESTIMATED OCCUPATIONAl RADIATION DOSE BY WORK FUNCTIONS FOR UNIT 2 Function Routine Operations and Surveillance Non-routine Operations and Surveillance Routine Maintenance Waste Processing Refueling In."ervice Inspection Special Maintenance Total Annual Dose Man-Rem Yr 56.0
- 32. 0 191.0
.54.0 2~.0 107.0
$5.0 5ZB.O Percentage of Total Dose lo. 4 9b.2
- 10. P.
Qo 5 l2. 3 100.0 1.15 1.16 1.21 1.22 1.24
- 1. 26-
- 1. 28
- 1. 30 1.32 1.34 Amendment
- 1. of 1
chl217718f-14co 09/05/84 112
Nine Mile Point Unit FSAR TABLE 12.4-5 OCCUPATIONAL DOSE ESTIMATES DURING ROUTINE OPERATIOIIS AND SURVEILLANCE
~Ae Y
Avg.
Dose Rate mrem hr Exposure
~F'me er No.
ol'orkers
~Fre cere t Dose
~mrem Yr
~Operal.ions Surveillance Reactor Building Turbine Building 1.3 4.0 Secur i'ty Surveys 1.0 Instrumentation and Controls 0.1 Chemistry Surveillance 0.5 1.5 2.0 5.5 0.50
- 6. 00 10 40 2/shift 2/shi ft Da i ly 1/hr 1/day 4.0 18.0
- 10. 0 4.0 9.0
- l. 18
- 1. 19
- 1. 20 1.22
- 1. 24
- 1. 26 Radiation Protection Surve i I lance 11.0
- 19. 0 1/wk Tote I 11.0 56.0
- 1. 28
- 1. 29 1.31 Amendment 1 of 1
ch1217718f-14cp 09/06/84 155
Nine Hile Point Uni fSAR TABLE 12.4-6 OCCUPATIOIIAL DOSE ESTIHATES DURING NON-ROUTINE OPERATIONS AND SURVEILLANCE A~ct vit Equipment Operat,ions RWCU System Condensate System RIIS System SFC System ECCS System SLS System Instrument Cal ibration Instrumentation and Cont,rois Radiation Honitors Linearity Checks Calibration AY9. Dose Ri1 to mrem hr 1.0
~
2 2 0.2 1.0 0.2 1.0 0.2 1.0 1.0 Exposure T~ime h r GO.O.
8.0 2;0 6.0
~
2.5 3.0 6.0
- 50. 0 225.0 Iio.
of Workers 2
2 2
2 2
2 40 1
2 rretruenc 1/yr 1/day 1/month 1/yr 1/month 1/month I/day 2/yr I/I.Syr Dose
~mrem r
- 0. 10 13.0 O.ill.
0 OI 0:01
- 0. 10 18.0
- 1. 18
- l. 19 1.20 1.21
- 1. 22
- 1. 23
- 1. 24
- 1. 26
- 1. 27 1.28
- 1. 30 1.31
- 1. 32 Tota I 3t.00
- 1. 34 Amendment 1 of 1
chl217718f-14cq 09/06/84 155
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Nine Mile Point Unit 2 F
TABLE 12.4-8 OCCUPATIONAL DOSE ESTIMATES DURING WASTE PROCESSING (RADWASTE OPERATIONS)
~AC ivi Y Avg.
Dose Rate
~@rem hr Operation of Liquid Radwaste 0.5 System Operation of Solid Radwaste 0.5 System Exposure T~ime hr
,6. 0 6.0 No.
of'orkers
~rre eche 1/shi ft 1/shift Dose jmmrem r
10.0
- 10. 0
- 1. 14 1:15
- 1. 17
- 1. 18
- 1. 20 1.21 DAW Compacting RadwasI,e Shipments DAW Shipments 3.5 1.0 5.0 6.0 6.0 16.0 2/day 1/wk 1/mont,h
~
Tota I 31.0 1.0 2.0 54.0
- 1. 23
- 1. 25 1.27 1.29 Amendment 1 of 1
ch1217718f-14cs 09/06/84 155
Nine Hile Poi Uni.t 2 FSAR TABLE 12.4-9 r
OCCUPATIONAL DOSE EST IHATES DURING REFUELING A~el v I Reactor Disassembly Reactor Assembly Fuel Unload Fuel Load Fuel Preparation Avg.
Dose Rate 12.0 12.0 2.0 2.0 2.0 Exposure
~Time hr 75.0 150.0 200.0 180.0 100. 0 No.
o I'orkers 10 10 Fre uenc r
1/1. 5 1/1'. 5
)/1. 5 1/1. 5 1/1. 5 Dose
~mr em r
6.0 12.0 2.0 2.0 1.0
- l. 15
- 1. 16
- 1. 18
- 1. 20
- 1. 22
- 1. 24
- 1. 26 Tota I 23.0
- 1. 28
- 1. 30 Amendment 1 of 1
ch1217718f-14ct 09/06/84 155
Nine Hile Poi Unit 2 FSAR TABLE 12.4-10 OCCUPATIONAL DOSE ESTIHATES DURING INSERVICE INSPECTION
~ee Avg. Dose Rate mrem hr Exposure
~Time hr No. of Horkers Fre uenc r
Dose
~ere r
- 1. 15
- 1. 16 Reactor Bui Iding -
100.00
~
Priaa ry'ontainment, 150 7
1/1. 5 70.0
- 1. 18
- 1. 19 Reactor Building-Secondary Containment, Turbine and Hiscel-laneous Buildings 10.00
- 10. 00 580 110 1/refue I ing 1/1. 5 31.0 6.0 1.21 1.22
- 1. 24 1.25 Tota I 107. 0 1:27
- 1. 29 Amendment 1 of 1
ch1217718 f-14cu 09/06/84 155
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Nine Hile Point Unit 2 FSAR TABLE 12.4-12
~PI an Dresden 1,2,3 OPERATIONAL HAN-REH PER YEAR fOR SELECTED BWR PLANTS
~174 I 997 I 991 j
~17 7
~178
~17 19¹
~18 1
~182 1,662 3,423 1,680 1,693 1
~ 529 1,800 2, 105 2.802 2,923
- l. 18 Hont.icello 349 1,353 263 1,000 Nine Hile Point 824 681 428 1,383 375 157 531 1,004 993 1.20 314 1,497 591 1,592 1,264 1.22 Quad Cities 1,2 482 1,618 1, 651 Peach Bottom 2,3 228 840 2,036 I ~ 317 1,388 2,302 2,506 1,977 1.24 1,031 1,618 2, 158 4,838 3, 146 3,757 1.26 Vermont, Yankee 216 153 411 258 339 1, 170 I, 138 731 205
- 1. 28 Pi Igrim 1
798 2,648 3,142 l.327 1.015, 3,626 1,836 1,539 1.30 Hillstone Point 1
1,430 2,022 1, 194 392 1; 239 1,793 2, 158 1,496 929
- 1. 32 Oyster Creek Brunswick 1,2 Brown ferry 1,2,3 fi tzpatrick 1,080 909 984 I, 140 1, 078 1, 614 1, 279 1,004 I, 792 2,602
, I, 1,66 I 859 3,870 2,638 3 '92 1.36 I 1,825 2,040 2,380 2,220 1.38 1,425 1, 190
- 1. 40 467 I ~ 733 917 865
- 1. 34 Avg mrem/un i t 594 878 Overall avg. of 948 mrem/year-unit Rel'erence:
NUREG-0713, Volume 4
784 974 686 872 1,419 1,183 1,140 1.42 1.44 i 1.46 Amendment 1 of 1
c h1 217718 f-14cw 09/06/84 155
Nine Mile Point Unit 2 TABLE 12.4-13 DISTRIBUTION OF ANNUAL MAN-REM BY WORK FUNCTIONS BASED ON OPERATING BWR DATA Work Function 1978 1979 1980 1981 1982 Ay~
- 1. 15 Reactor Operations and Surveillance Routine Maintenance Waste Processing Refueling Inservice Inspection Special Maintenance
- 5. 8.
4.3 2.0 4.4 2.6 7.3
- 34. 1 31
~ 2 12.3 13 '
43.2 39.3 7.6 7 '
42.8 42.2 3.1 11.0 5.2 2.5 3.3 3.7 38.1 33.1 9.1 10.0 1.17 1.18 4.3 4.2 1.26 44.0 36.1 1.28 33.7 40.2 1.20 6.2 6.1 1
~ 22 2.7 3.4 1.24
References:
UREG-0594, "Occupational Radiation Exposure at Commerical Nuclear Power Reactors, 1978,"
- November, 1979.
1.30 1.32
- l. 34 1.35 NUREG-071/,
Volume 1, "Occupational Radiation Commercial Nuclear Power Reactors;
- March, 1981.
Exposure at 1979,"
1.37 1.38 1.39 NUREG-0713, Volume 2, "Occupational Radiation Exposures of Commer al Nuclear Power Reactors, 1980, "
- December, 1981
~
NUREG-0713, Volume 3, "Occupational Radiation Exposure at Commercial Nuclear Power Reactors, 1981,"
- November, 1982.
NUREG-0713, Volume 4, "Occupational Radiation Exposure at Commercial Nuclear Power Reactors, 1982,"
- December, 1983.
1
~ 41 1.42 1.43 1.45 1.46 1.47
- 1. 49 1.50 1.51 Amendment 1 of 1
chl217718f-14cx 09/06/84 155
'y
Nine Mile Point Unit 2 FSAR Waste gas includes headers and cover gas system outside of containment in addition to decay or storage system.
Include a list of systems containing radioactive materials that are excluded from the program and provide justification for exclusion.
Testing of gaseous systems should include helium leak detec-tion or equivalent testing methods.
A program should be considered to reduce leakage potential release paths due to design and operator deficiencies as discussed in our letter dated October 17,
- 1979, to all operating nuclear power 'plants regarding North Anna and related incidents.
Nine Mile Point Unit 2 Position A program.has been developed to monitor leakage from systems outside the containment which could be used to transport highly radiaactive fluids in a
post-accident condition.
This program inct udes the following features:
a.
2.
The, implementation of a periodic visual inspection program consisting of a combination of general in-spections and detailed system walkdown of liquid systems.
These inspections shall be performed on accessible portions
" of applicable systems during system operational testing or by evaluation of leakage at lower pressures during operation.
Systems containing gases are to be tested by use of tracer gases (helium, freon or DOP) by pressure decay testing or by metered makeup tests.
3.
An aggressive maintenance program will be used to assign high priorities to leakage-related Main-tenance Work Requests (MWRs).
5.
Preparation of systems list, identifying specific methods used to test systems, the system
- involved, and frequency of testing.
Records shall be maintained on the tests and in-spections performed and leakage related MWRs.
These records shall be used to identify chronic and generic leakage problems in order to implement modifications
. and/or corrective maintenance measures to keep leakage as low as practical.
1.10-125
tg
Ni,ne M~le Point Unit 2 FSAR prior+ wlLeak a~@'pgImabg Qr mmes These mea ures will be implementedh, pri8r to full power operation.
NMPC will submit, a report to the NRC Staff of all recorded leakage and all preventive main-tenance performed as a direct result of the evaluation of this leakage.
The report will also identify general leakage criteria to be applied during the first fuel cycle as the basis for instituting a
corrective action in the form of preventative maintenance.
Prior to the start of the second fuel cycle; NMPC will revise the general criteria as neces-sary based on the experience gained during the Unit 2 first fuel cycle.
The revised criteria shall then be used as the basis for long-term leakage monitoring activity at Unit 2.
~ ~
~ 44
- 1.10-126
Nine Mile Point Unit'. 2 FSAR 2.
Adequate NPSH to the RHR pumps is provided with 50 percent of the, strainer area clogged.
3.
Strainers are designed to withstand any loads during suppression pool transients, such as temperature,
- pressure, and water level.
~ Strainers are designed to withstand a
pressure differential of 25 psi.
All strainers are seismically qualified.
Insulation Types of insulation used for piping and equipment within the drywell and suppression chamber
'are discussed in the following paragraphs.
For piping and equipment located within the drywell, that require insulation to minimize heat loss, primarily metal-reflective-type insulation is used.
Metal-reflective insulation is an all-metal construction-type insulation that has a
stainless steel.
inside and outside jacket which encapsulates multiple layers of stainless steel insulation material.
Metal-reflective insulation is installed in sections with overlapping edges and quick-release latches with keepers.
Two other types of insulation are used inside the drywell for special and limited application:
Min-k and Temp-Mat insulation.
Min-k is a powder-type insulation used where space is limited'nd is encapsulated in stainless steel so as to be watertight.
Temp-Mat is a borated, spun glass, blanket-type insulation used where it is necessary to lower the neutron flux (i.e.,
at the primary shield wall penetration) and is also encapsulated in stainless steel.
(See Table 6.2-64)
No anti-sweat insulations are used within the primary containment.
The mechanism for transpor't of any insulation debris from the drywell into the suppression pool following an accident involves a
series of unlikely occurrences, as discussed in the following paragraphs.
In the event of a postulated pipe break, some insulation in the immediate vicinity of the break could possibly be removed by direct jet impingement.
Since the insulation is fabricated and installed in overlapping
- sections, only sections in the immediate vicinity of the break would likely be affected.
The stainless steel
. jacket minimizes the 6.2-44
TABLE 6.2-64 MATERIAL Temp-Mat VOLUME 122.25 ft. 3 255 MARGIN 153 ft. 3 Min-K 91.5 ft. 3 115 ft. 3
Nine Nile Point Unit 2 FSAR Adjustments are then made to compensate for differences in test (air) and LOCA environment (steam/air) mediums and standard (scf) leak rates converted to drywell temperature and pressure conditions.
The instantaneous leakage rate is then plotted against time for 30 days.
The maximum bypass leakage rate is determined by taking the area under this plot.
The maximum bypass leakage rate contributed by the isolation valve is 268.93 cf distributed over 30 days.
The calculated maximum bypass leakage volume following a design basis LOCA is 2,269 cf over 30 days.
The individual and total line leakage volume for selected periods during the 30 days are given in Table 6.2-55.
6.2.3.3 Design Evaluation 6.2.3.3.1 LOCA Temperature and Pressure Transient During normal plant operation the reactor building and auxiliary bays are maintained at a
negative pre'ssure relative to atmosphere of 0.25 in W.G..
by the reactor building ventilation system described in Section 9. 4.
In the event of a LOCA, the reactor building ventilation system is isolated and the standby gas treatment system (SGTS) is initiated upon receipt of any of the three signals listed in Section 6.2.3.2.2.
Details of the SOTS are provided in Section 6.5.1.
The reactor building and auxiliary bays-are considered one volume which is at a uniform pressure.
6.2.3.3.1.1 Summary and Conclusions The post-lOCA transient response of the reactor building and auxiliary bays atmosphere has been analyzed for a
duration of 96 hr, as shown on Figure 6.2-77.
The temperature and pressure stabilize prior to 96 hr.
The characteristics of the. transient responses may be summarized as follows:
1.
The transient responses of the reactor building and auxiliary bays pressure and temperature are shown on Figures 6'.2-76 and 6.2-77,
. respectively..
2.
The SGTS centrifugal exhaust fan characteristics are shown on. Figure 6.2-78.
Amendment 4
6.2-57
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Nine Mile Point Unit 2 FSAR 5.
The capacity of one SGTS
- train, 3600 cfm, is adequate to restore and maintain the reactor building and auxiliary bays pressure at or below
-0.25 in W.G., relative to atmosphere after a
?OCA, as shown on Figure 6.2-76.
6.
The period during which the pressure profile is greater than'..-0=.25 in W.G.
is indicated on Figure 6.2-76 and lasts approximately 75 sec.
The analytical results-,
based on the assumptions in Section 6.2.3.3.1.3, show that the SGTS will accomplish its design objective of maintaining a pressure equal to or below
-0.25 in'.G.
within the reactor building and auxiliary bays following a fOCA.
6.2 '.3.1.2 Calculation Approach The analysis was performed assuming that the reactor building and auxiliary bays are one large constant volume.
One SGTS filter train was considered in operation.
The inleakage was assumed to be 100 percent of the reactor building and auxiliary bays volume per day at the design outside air temperature of 93 F.
The heat transfer between the outside environment, and the reactor building and auxiliary bays was considered since this results in a
net positive heat gain to the reactor building and auxiliary bays.
Amendment 4
6.2-57a September 1983
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Nine Mile Point Unit 2 FSAR 6.2.3.3.1.3 Assumptions Some of the assumptions applied to this analysis were:
A LOCA and loss of offsite power are assumed to occur simultaneously.
Emergency power is assumed to be
'upplied by two of the three diesel generators, considering the failure of either the Division I or Division II diesel generator.
2.
Nonadiabatic boundary conditions are assumed for the surface of the reactor building and auxiliary bays structure exposed to the outside environment.
Nonadiabatic boundary conditions result in a
net positive heat gain which is more conservative than adiabatic boundary conditions.
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6.2-57b
Nine Mile Point Unit 2 FSAR 6.
The heat gain from the spent fuel pool is based on the maximum normal spent fuel pool temperature of 125 F.
The reactor building is assumed to be sufficiently leaktight to limit the inleakage to 100 percent of the reactor building and auxiliary bays volume per day with a
-0.25-in water differential pressure under neutral wind loading conditions.
8.
For mechanical equipment and its associated
- piping, which operates only on Division I or Division II powe'r, it is assumed that there will be no heat gain when the equipment is not energized as
-a result of the failure of the respective division diesel generator.
9.
During a large break LOCA, it is assumed that there will be no flow or heat gain to the suppression pool through the high pressure core spray pump test return line.
10.
It is assumed that the recirculation loop of the SGTS does not operate during the analysis.
11.
The compressive effect of primary containment expansion is assumed to be insignificant.
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6.2.3.4 Test and Inspection Tests and inspections of the reactor building ventilation system and the SGTS will be performed prior to initial fuel load and periodically 'hereafter in accordance with technical specification requirements.
6.2.3.5 Instrumentation Requirements A
reactor building negative air pressure of 0.25 in W.G. is automatically maintained under normal operating conditions by the reactor building ventilation system.
- Normally, modulating air dampers automatically recirculate supply air to maintain negative pressure in the reactor building.
During accident conditions (LOCA), isolation dampers in the air supply and air exhaust ducts will close automatically; Amendment 4
6.2-57c September 1983
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182b The a
licant has committed to erform re-o erational and periodic tests of the standby gas treatment system to verify that each train will attain 3,500 fthm/min rated flow.and will aeet its design oh)ective.
The staff will require
.that, as art of this test acce tance criteria, the applicant make provisions to determine the econde containment de reeeurizetion time, serif the nleaka e rate of 3 160 fthm min, the uniformit of ne ative ressure throughout
,the seconda containmen,.
and he otential for exfiltration.
The staff will report on this in a supplement to this report.
a)
See Revised Test Abstract 14.2-77.
b)
See Revised Test Abstract 14.2-77 c)
The Emergency Recirculation System ensures mixing throughout the reactor building atmosphere.
d)
See Section 6.2.3.3
, Nine Mile Point Unit 2 FSAR TABLE 14.2-77 STANDBY GAS TREATMENT AND SECONDARY CONTAINMENT LEAKAGE TEST System 61 Preo erational Test N2-POT-61B Test Ob'ectives 1.
To demonstrate the reliable operation of the standby gas treatment system and components.
2
~
To verify that the standby gas treatment system can maintain the propex reactor building pressure and that reactor building leakage xate is within design limits.
12 Safet Precaution Follow all NMPC safety rules and proper procedures during testing.
Prere isites l.
All applicable preliminary tests are completed and approved.
2.
All applicabl'e motor control centexs to supply electric power to motors, control circuits, and instrumentation are available.
3.
All valve lineups are completed.
4.
Reactor building ventilation system is operable",
and all reactor building doors and hatches are closed.
12 Test Procedure 1.
The test procedure will verify that the two gas treatment filter trains operate according to design specifications under normal and transient conditions.
2.
Various system-auto initiations will be demonstrated.
3.
System annunciators, control instrumentation, and intex.locks will be tested.
12 4.
Standby gas treatment fan operation will be verified.
Amendment 12 1 of 2 June 1984
Nine Mile Point Unit 2 FSAR TABLE 14.2-77 (Cont) 5..
The test will verify that the SGTS will accomplish its design objective of reestablishing the Reactor Building pressure equal to our below -0.25 in W.G. within the required time interval.
6.
With the standby gas treatment system in operation and all doors and hatches controlled in the closed position, secondary containment leakage rate will be verified as within allowable limits.
Acce tance Criteria 1.
Each standby 'as treatment system train and its associated equipment,
- valves, motors, filters, etc, will function as designed according to SWEC logic drawings.
2.
System interlocks, control instrumentation, and annunciator's function as designed according to SWEC design drawings.
3.
Reactor building ventilation system isolation functions as designed according to system logic drawings.
4.
Each standby gas treatment system train reactor building pressure equal to or below 5.
The reactor building leakage rate is not 3,160 cfm.
can maintain
-0.25 wg.
greater than 6.
The secondary containment drawdown time to -0.25 in. W.G. is less than 90 seconds.
Amendment 12 2 of 2 June 1984
Nine Mile Point Unit 2 FSAR QUESTION F480.22 (6.2.3)
Bypass leakage is defined as that leakage from the primary containment which can circumvent the secondary containment boundary and escape directly to the environment, i.e, bypassing the leakage collection and filtration system of the secondary containment.
FSAR Table 6.5-56 indicates that most piping lines are not potential bypass path.
List the lines so designated and indicate why they are not bypass leakage paths.
Systems lines may be excluded from con-sideration as potential bypass paths for reasons such as:
the lines terminate in the secondary containment, an air 'r water sealing system is provided to process or eliminate
- leakage, or a closed system is proposed for the leakage boundary.
If a
closed system is proposed as the leakage boundary to preclude bypass leakage verify that the fol-lowing provisions of SRP 6.2.3 are satisfied.
The system should:
a ~
Either (1) not directly communicate with the con-tainment atmosphere, or (2) not directly com-municate with the environment, following
" a loss-of-coolant accident.
b.
Be designed in accordance with Quality Group B standards, as defined by Regulatory Guide 1.2'6.
(Systems designed to Quality Group C or D standards that qualify as closed systems to preclude.
bypass leakage will be considered on a
case-by-case basis.)
c.
Meet seismic Category I design requirements.
d.
Be designed to at least the primary containment pressure and temperature design conditions.
e.
Be designed for protection against pipe
- whip, missiles, and jet forces in a
manner similar to that for engineered safety features.
f.
Be tested for leakage, unless it can be shown that during normal plant operations the system integrity is maintained.:
Specify the estimated bypass leakage for penetrations which must be-considered as bypass paths.
RESPONSE
~'4lSRP ~c<F
>Amendment 5
QSR F480.22-1 October 1983
Nine Mile Point Unit. 2 FSAR g<<ISED TABLE 6.2-56 (Cont)
KEY TO ISOLATION SIGNAIS:
A = Low reactor vessel water level 3
B = I,ow reactor vessel water level 2
C = High main steam line radiation D = High main steam line flow E = High main steam line tunnel area ambient temperature F = High drywell pressure G = Low reactor vessel water level 2
or high drywell pressure J = High reactor water cleanup system equipment area differential and ambient temperatures K = Reactor core isolation cooling high pipe routing and equipment area temperature, low steam supply pressure.
High steam line differential
- pressure, high turbine exhaust diaphragm pressure L = High reactor vessel pressure M = High residual heat removal system ecpxipment area differential and ambient temperatures P
= Low main steam line turbine inlet pressure R = Low main condenser vacuum S
= Standby licpxid control system actuated T = High main steam line tunnel differential and ambient temperatures U = High reac or water cleanup svstem different'l flow W = High reactor water cleanup system nonregenerative heat exchanger outlet temperature LC = Locked closed RM = Remote manual switch from control room Z5 17 of~
Nine Mile Point Unit 2 FSAR TABLE 6.2-56 (Cont)
LMC = Local manual control, locked closed, position indica-tion in control room
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1M'1C Rgb'~ P y1P NOTES:
Tyoe C testing is discussed on Figure 6.2-70 which shows the isolation valve arrangement.
Norma3.
status position of valve (open or closed) is the position during norma3.
power operation of the reactor (see Normal Position column).
Primary containment and reactor vessel isolation signals are indicated by letters.
isolation signals
- generated by the individual system process control signals or for remote manual c3.osure based on information available to the ope ator are discussed in the referenced notes in the isolation Signa3.
column.
The specified closure rates are as required for containment isolation or system operation, whichever is less.
Reported times are in seconds.
The standard minimum closing rate is 12 in/min of nominal valve diameter for gate valves and 4 in/min of valve stem travel for globe valves.
For example, a
'2-in gaze valve will close 'n i min.
Ac motor-operated valves required for isolat'on functions are powered from the ac standby power buses.
Dc-operated isolation valves are powered from safety related station batteries.
A main steam isolation valve requires that one spring latch be released to close the valve.
Two springs are provided for redundancy.
The valves are designed to fully close within 3 to 5 sec.
18 of
Nine Mile Point Unit 2 FSAR TABIE 6.2-56 (Cont)
<9) All isolation valves are Category I.
' 'All motor-operated isolation valves remain in the as-is position upon failure of valve power (FAI =.Fail as is).
All air-operated valves close on mot've air failure in the safe position.
'Testable check valves are designed for remote opening with zero differential pressure across the valve seat.
The valves will close on reverse flow even though the test switches may be positioned for open.
The valves open when pump pressure exceeds reactor pressure even though the test switch may be positioned for close.
'These valves are the ECCS and drywell spray suction and discharge isolation valves.
ECCS operation is essential during the LOCA period; therefore, there are no automatic isolation signals.
A high level alarm in the appropriate reactor building sump indicates excessive ECCS leakage into the secondary containment.
'Suppression eeoc s} mY valves have interlocks that allow them to be manually reopened after automatic closure.
This setup permits suppression pool spray, for high drywell pressure conditions When automatic signals are not present, ttiese valves may be opened for test or operating convenience.
~'Due to redundancy within -the
- ECCS, some subsystems may be secured during the long-term cooling period.
In
- spray, suppression chamber
- spray, suppression pool cooling) which the operator may select during the 30-day post-LOCA period.
'The RCIC steam exhaust
- valve, 2ICS*MOV122, is normally open at all times.
Should a leak occur, it would be detected and alarmed by the RCIC room nigh temperature leak detection system.
7'C 'erion 55 concerns lines of the reactor coolant pressure boundary (RCPB) that penetrate the primary reactor containment.
The CRD insert and withdraw lines are not part of the RCPB.
The classification
'of the 19 of~
g Qv]gaul Nine Mile Point Unit 2 FSAR TABIE 6.2-56 (Cont) inse t 'nd withdraw lines is Quality Group B, and therefore they are designed in accordance with ASME Section III, Safety Class 2.
The basis to which the CRD 'nes are designed is commensurate with the safety importance of isolating these lines.
Since these lines are vital to the scram function, their operability is of utmost concern.
In the desi'gn of this
- system, it has been accepted practice to omit automatic valves for isolation purposes as this introduces a possible failure mechanism.
As a
means of providing pos'ive actuation, manual shutoff valves are used.
In the event of a break on these
- lines, the manual valves may be closed to ensure iso3.ation.
In
- addition, a ball check valve located in the insert line inside the CRD is designed to automatically seal this line in the event of a break.
~ 'The operator' indication that remote-manual closure of the TIP shear valves is required is failure of the TIP
. ball valves to close.
~'Since the traversing incore probe (TIP) system lines do not communicate freely with the containment atmosphere or the reactor coolant, General Design Criteria 55 and 56 are not directly applicable to this specific class of lines.
The basis to which these lines are designed is more c3.osely described by Criterion 57, which states in effect that isolation capability of a system should be commensurate with tne safety importance of that iso3.ation.
2'urthermore, even though the failure of the TIP system lines presents no safety consideration, the TIP system has redundant iso iation capabilities.
The safety features were reviewed by the NRC for BWR/4 (Duane Arnold),
BWR/5 (Nine M'le Poin-Unit 2) and BWR/6 (GESTAR II), and it was concluded tha the design of the containment isolation system meets the objectives and intent of the general design criteria.
Isolat'on is accomplished by a seismically qualified, solenoid-operated ball valve that is normally closed.
To ensure isolation capability, an explosive shear valve is installed '
each l'ne.
Upon receipt of a
signal (manua3.ly initiated by the operator),
this explosive valve will shear the TIP cable and seal the gu'de ube.
20 of~
Nine Mile Point Unit 2 FSAR TABLE 6.2-56 (Cont)
When the TIP system cable is inserted, the ball valve of the selected tube opens automatically so that the pzobe and cable can advance.
A maximum of five valves can be opened at any one time to conduct calibration, and any one guide tube is used, at most, a few hours per year.
If closure of the line is required during calibration, a
signal causes a cable to be retracted and the ball valve to close automatically after completion of cable withdrawal.
If a TIP cable fails to withdraw or a
ball valve fails= to
- close, the explosive shear valve is actuated.
The ball valve position is indicated in the control room.
The Unit 2 TIP system design specifications require that the maximum leakage rate of the ball and shear valves be
~
in accordance with the Manufacturer's Standardization Society (hydrostatic testing of valves).
The TIP isolation valve and the shear valve both have a
leak integrity requirement of 10
~ atm cc/sec for air-water combination and water alone.
This leakage rate represents less than 10
~ cc/sec 'of fluid at the following conditions:
Air-water combinations:
0-125 psig and 300 F
Water:
1,250 psig and
<450 F
As stated
- above, the penetration is automatically closed following use.
During normal operation the penetration will be open approximately 8 hz/month to obtain TIP infozmation.
If a failure occu red, such as inability to withdraw the TIP
- cable, the shear valve could be closed to isolate the penetrations.
Installation requirements are that the guide tube/penetration flange/bali and shear valve composite assembly not leak at a
rate greater than 10 'tm cc/sec at 125 psig.
Further leak est'ng of the shear valves not recommended since destructive testing would be required..
K,eak testing of the ball valves also is not recommended since the guide tube terminates in a
sealed indexer hous'g that is kept under a positive pressure by a
nitrogen purge.
The purge makeup is indicative of system leakage.
Note that the TIP ball valve is normally closed and thus is a
part of the leakage 21 of~
z,s
Nine Mile Point Unit 2 FSAR TABLE 6.2-56 (Cont) barrier being monitored.
Consequently, the personnel exposure required to conduct Type C
tests from inside the containment is not warranted.
'~4'Removable spool piece that is removed during normal operation; it is installed when the plant is down and fi e protection is needed inside the primary containment.
Air-operated valve's 104 and 106 are manually operated before personnel entry into the primary containment.
Line length is given for the most remote valve.
'ystem isolation valves are normally closed.
The system is placed in operation only if the hydrogen monitors detect hydrogen buildup after a
LOCA.
The operator has flow indication, in the main control
- room, of gas leaving and entering the containment.
Should these flows vary significantly from one anothe, it would be detected in the main control room and the process loop in service could be shut down.
The valve is open only during steam condensing mode.
Valve position is indicated in the maj.n control room to provide the operator confirmation of valve status.
'~~'This line consists of the following inputs from these valves:
2RHS*SV34A and 2RHS~SV62A - steam condensing line safety valves.
2RHS*RV56A -
RHR heat exchanger shell side elief valve.
2RHS*MOV26A and 2RHS*MOV27A -
RHR neat excnanger vent line isolation valves.
2RHS~V20 and 2RHS*V19 - vacuum breaker l'ne.
The valve is open only during steam condensing mode.
Valve position is indicated in the main control room to provide the operator confirmation of valve status.
22 of Qs
88gjsz ~
Nine Mile Point Unit 2 FSAR TABIE 6.2-56 (Cont)
This line consists of the following inputs from these valves:
2RHS*SV34B and 2RHS*SV62B - steam condensing line safety valves.
2RHS*RV56B -
RHR heat exchanger shell side relief valve.
2RHS*MOV26B and 2RHS*MOV27B -
RHR heat exchanger vent line isolation valves 2RHS*V117 and 2HS*V118 - RCIC vacuum breaker line.
The valve is open only during steam condensing mode.
Valve position is indicated in the main control room to provide the operator confirmation of valve status.
'Normally closed.
Opened
,only when testing wetwell to drywell vacuum breakers.
'~~'Penetrations Z-99A,B,C,D, and Z-100A,B,C,D contain lines for the hydraulic control of the reactor recirculation flow control 'valve.
These lines contain hydraulic fluid used to position the reactor recirculation flow control valve.
These lines inside the containment are Category I and Quality Group B.
They have failed-closed automatic isolation valves outside the conta'ment which receive an automatic isolation signal on high drywell pressure.
These lines meet the requirement of General Design Criterion 57 and therefore require only single automatic
'olation valves outside the containment.
They also meet the requirement of Standard Review Plan
- 6. 2. 4.
They are designed to Category I, Code Group B,
and the following cr'eria:
C.
Do not communicate with either the reactor coolant system or the containment atmospnere.
Are protected against m'ssiles and p'pe w¹p.
Will withstand temperatures at least equal to the containment design temperature.
23 of~
Nine Mile Point Unit 2 FSAR TABTE 6.2-56 (Cont)
'~"'This line consists of the following inputs from these valves:
2RHS*SV34B and 2RHS*SV62B - steam condensing line safety valves.
1.17 1.18 2RHS*RV56B -
RHR heat exchanger shell side relief valve.
1.22 1.23 2RHS*MOV26B and 2RHS*MOV27B RHR heat.
exchanger vent line isolation valves 1.27 1.28 2RHS*V117 and 2HS*V118 - RCIC vacuum breaker
, line.
~
1.32 1.33 The valve is open only during steam condensing mode.
Valve position is indicated in the main control room to provide the operator confirmation of valve status.
'Normally closed.
Opened only when testing wetwell to drywell vacuum breakers.
'~~'Penetrations Z-99A,B,C,D, and Z-100A,B,C,D contain lines for the hydraulic control of the reactor recirculation flow control valve.
These lines contain-hydraulic fluid used to position the reactor recirculation flow control valve.
1.37 1.38
- 1. 42 1.45 1.47 Amendment 8
23 of~
January 1984 ch1217718f-8gw 01/17/84 105
Nine Mile Point Unit 2 CESAR TABIE 6.2-56 (Cont) d.
Will withstand the external pressure from the containment structural acceptance test.
e.
Will withstand the IOCA transient and environment.
Even if the failed-closed valve were to not shut there would be no leakage of containment atmosphere through the hydrau'ic control lines since the piping inside the prima v containment would remain intact.
There are no active component failures that would compromise tne integrity of the closed system inside the primary
, containment.
Integr'ty of the closed system inside the primary containment is, essentially, constantly monitored since the system is under a constant operating pressure of 1,800 psig.
Any leakage through this system would be noticed because operation would be erratic and because of indications provided on the HCU.
In
- addition, in order to perform Type C tests on these
- lines, the system would have to be disabled and drained of hydraulic fluid.
This is considered to be detrimental to the proper operation of the system since possible damage could occur in establishing the test condition or restoring the system to normal.
These lines and associated isolation valves should therefore be considered to be exempt from containment testing.
'nstrument lines that penetrate primary containment conform to Regulatory Guide 1.11.
The lines that connect to the reactor pressure boundary
-include a
restricting orifice inside containment, are Category I, and terminate in instruments tha" are Category I.
The instrument lines also include manual isolation valves and excess flow check valves or equivalent.
These penetrations will not be Type C
tested since the integrity of the lines is continuously demonstrated during plant operations where subject to reactor opera ing pressure.
In addition, all lines are subject to the Tvpe A
test pressu e
on a
regular interval.
Leaktight integrity is a'so verif'ed with complet'on of funct'onal and cal'brat'on surve'llance ac"ivities as well as by visual obse vations dur'g operator tours.
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Nine Mile Pbint Unit 2 FSAR QUESTION F480.29 (6.2.4)
Describe the provisions to insure that debris will not become entrained in the purge valves and prevent their closure.
Guidance is provided below which, if followed, would represent an acceptable debris screen design:
1.12 a) b)
The debris screen should be seismic Category I and installed.typically about one pipe diameter away from the inner side of the inboard isolation valve.
The piping between the debris screen and the valve should also be seismic Category I design.
1.17 1.18 1.19 1.20 1.21 c)
The debris screen should be designed to withstand the COCA diTferential pressure.
d)
The debris screen openings should be about 2 inches by 1 3/16 inches.
A suggested debris screen design is enclosed as Figure 1.
RESPONSE
- 1. 22
- 1. 23 1.24 1.25 1.27 1.29 Debris
- screens, in general accordance with the design referenced above as Figure 1,
are provided.
See revised Sections 6.2.4.3.2 and 9.2.4.
1.30 1.31 1.32 Amendment Q6cR F480.29-1 ch12177f qr14f 08/28/84
Nine Mile Point Unit 2 FSAR would be insignificant.
Suppression pool makeup during normal plant conditions is from the condensate water storage tank.
- 1. 12 1.13 The elevations of the ECCS pump suction centerlines and the suppression pool minimum drawdown level are 195 '0" and 197'-8", respectively.
Influent and Effluent Lines from Dx well and Su ression Chamber Free Volume 1.14 1.15 1.16 1.18 Primar Cont:ainment:
Pur e'Lines The dxywell and suppression chamber purge lines have isolation capabj.lities commensurate with the impoxtance to safety of iqolating these lines.
Each line has two normally ~osed/fai,l closed valves - one located inside (nitrogen operated) and one located outside (air operated) the primary containment.
The inboard end of each 12-in and 14-in valve located inside the primary containment is provided with.a QA Category I debris screen to prevent entrainment of foreign matter in the valve seat.
The isolation valves are interlocked to preclude opening of the valves while a primary containment isolation signal exists (Table 6.2-56).
1.20 1.21 1.22 1.2 1.25 1.26 1.27 1.2 1.30 1.31 2.
Primar Containment Atmos here Monitorin S stem Sam lin Lines The primary containment atmosphere monitoring
. system consists of radiation and hydrogen/oxygen moni toring lines.
Each
- line, suction and discharge, penetrates the primary containment and continuously monitors the radiation level and hydrogen/oxygen concentration during normal operation.
These lines are equipped with two solenoid-operated isolation valves, one inside the primary containment and the other
- outside, located as close as possible to the primary containment.
The hydrogen/oxygen monitoring lines are also used to continuously monitor the pximary containment air during the post-LOCA period.
Each isolation valve receives isolation signals.
The isolation valves for hydrogen/oxygen monitoring lines are provided with individual keylock switches to override the isolation signal and initiate-system operation, during the post-LOCA period.
1.32 1.33 1.34 1.35 1.36 1.37 1.38 1.39 1.40 1.42 1.43 1.44 Amendment 6.2-68a ch1217718f-14bb 08/28/84 114
Nine Mile Point Unit 2 FSAR space and stairways into the return air system located at each floor level.
The emergency recirculation system ensures mixing throughout the reactor building atmosphere, including the spent fuel pool area.
- 1. 13 1.14 1.15 The intake duct connection for the SOTS (Section 6.5.1) is taken at the discharge side of the emergency recirculation unit cooler to maintain the reactor building at a negative pressure.
Unit space coolers with sufficient capacity to satisfy the cooling requirements of the emergency safeguard equipment provide cooling to handle the heat gain load of the respective safeguard equipment.
Cooling for general areas is provided by unit ~ace coolers.
HVAC equipment and components that operate following a TOCA are designed to Category I
and Safety Class 2
and 3
criteria.
Equipment motors and controls in the safety-related portion of the system are supplied from their respective independent emergency power sources and have sufficient redundancy to satisfy the single-failure criterion.
9.4.2.3 Safety Evaluation The safety features of the reactor building HVAC system are as follows:
1.23 1.24 1.25 1.26 1.27 1.28 1.30 1.32 1.33 1.34 1
~ 35 1.37 1.39 All safety-related components are designed to Safety Class 2
and 3
criteria and Category I requirements.
Safety-related components are located so that failure of a
portion of other nonessential systems does not prevent operation of any safety-related system.
1.42 1.43 1.44 1.45
- 1. 46 2.
Safety-related components have sufficient redundancy to meet the single active failure criteria.
The Failure Modes and Effects Analysis (FMEA) of the reactor building HVAC system is provided in the FSAR FMEA report.
1.47 1.48
- 1. 49 1.50 3.
Redundant isolation valves in each line penetrating the primary containment are in accordance with ASME Section III.
The piping between the isolation valves is Safety Class 2 and both the valves and piping are designed to Category I.
All other system piping is seismically supported.
The inboard end of each 12-in and 14-in CPS isolation valve located inside the primary containment is 1.51 1.52 1.53 1.54 1.55 1.56 1.57 Amendment
- 9. 4-26 chl217718f-14bu 08/28/84 112
Nine Nile Point Unit 2 FSAR provided with a
QA Category I debris screen to 1.58 prevent entrainment of foreign matter in the valve 2.1 seat.
All primary containment penetrations associated 2.2 with the reactor building HUAC have redundant Amendment 9.4-26a ch12 177 18f-14bu 08/28/84 112
Nine Mile Point Unit' FSAR QUESTION F480.24 Indicate what mechanisms are available to control drywell and wetwell pressure perturbations during normal operation.
Would this system be open to the SGTS in the event of a LOCA? If so, show that the SGTS is capable of withstanding the LOCA pressure and the system, filters are capable of radionuclide exposure and will still perform its intended function post-lOCA.
RESPONSE
DEE- <<<~~~~ ~~ioA 9.9.2..Z.Z Amendment 11 QBR F480.24-1 June 1984
Nine Mile Point Unit 2 FSAR QUESTION F480.38 (6.2.4)
The FSAR does not specifically identify the extent of drywell-suppression chamber purging that may be necessary during normal plant operations.
Discuss the manner in which Nine Mile Point 2 conforms to the requirements of Branch Technical Position CSB 6-4.
Indicate how small pressure perturbations will be accommodated in the containment.
RESPONSE
s~ REU(sFP 5~TLD&$ ~. z. 5.z.- 9 ~MD p.q.p.z.p, Amendment 11 QEcR F480.38-1 June 1984.
Nine Mile Point Unit 2 FSAR accident.
.Once placed in operation, the system continues to operate until it is manually shut down when an adequate margin below the hydrogen or oxygen. concentration design limit is reached.
The operation of the system can be tested from the control room.
The test consists of energizing the blower and heaters and observing system operation to see if components are performing properly.
Flow and pressure measurement devices are periodically calibrated.
Cooling water required for operation of the system is taken from the service water system.
.The cooling water is used to cool the water vapor and the residual gases leaving the recombiner prior to returning them to the primary containment.
During normal operation the recombiner system will be maintained in an inerted condition with nitrogen, ready for immediate startup.
6.2.5.2.3 Primary Containment Nitrogen Enerting System Oxygen control within primary containment during normal plant operation is achieved 'y means of the nitrogen inerting system.
During normal plant operation, oxygen concentration is maintained at or below 4 volume percent using this system.
The system is designed to supply nitrogen to the primary containment. for initial inerting and for makeup during normal operation.
6.2.5.2.4 Primary Containment Purge Primary containment purge capability is provided in accordance with Regulatory Guide 1.7 and as an aid in cleanup following an accident.
This function is fulfilled by the combined operation of the primary containment purge system (cps) A4D~ 57+AQQQ 645 w~v~T'V~~N +~+++b.
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Nine Mile Point Unit 2 FSAR DBE.
The system is designed to nonnuclear safety standards and is not required for safe shutdown of the plant.
9.4.2.1.2
.Primary Containment.Purge Power Generation Desi n Basis 1.
Provide sufficient p'urging capability for the primary containment to permit entry of personnel within 16 hr of a reactor cold shutdown.
2.
.3.
Provi e
a means o
maintain~cQ the primary containment at positive pressure ~'uring normal operation so that any leakage can be monitored.
~
wi+n niHOp,h Provide a,
backup system to the redundant hydrogen recombiners for the dilution of hydrogen following a
loss-of-coolant.
accident (LOCA).
The hydrogen recombiners are described in Section 6.2.5.
Safet Desi n Basis Provide seismically qualified piping and valves to protect adjacent safety-related equipment in the event of a
DBE.
The system is designed to nonnuc'lear safety standards and is not required for safe shutdown of the plant.
9.4.2.1.3 All Other Reactor Building Areas I
Power Generation Desi n BaMs Provide an environment that ensures habitability of the areas served and optimum performance of equipment, within the temperature limits shown in Table 9.4-1.
2.
For normal plant operation, provide a once-through ventilation
- system, utilizing outdoor air with controlled discharge of exhaust air to the atmosphere.
3.
Exhaust more air from the reactor building than is being supplied, thereby maintaining the area at a
negative pressure to inhibit the exfiltration of airborne contaminants.
Provide the capability to clean up the reactor pressure vessel (RPV) head during the refueling operation with the help of the reactor head evacuation filter assembly.
9.4-21
Nine Mile Point Unit 2 FSAR 9.4.2.2.2 Primary Containment Purge The primary containment purge system is shown schematically on Figure 9.4-B.
The system is divided into two su sys em' first subsystem purges the primary containment and consists of one 100-percent capacity centrifugal
- fan, piping,
- valves, controls, and accessories.
The fan draws makeup air from the reactor building ventilation syst: em and dischargeh through pipe ducts to the primary containment.
The SGTS (Section 6.5.1) takes suction through pipe ducts to exhaust the primary containment.
This subsystem also provides a
connection for a
portable compressor that performs the integrated leak rate test.
k The second subsystem pressurizes the primary containment.
It consists of piping, valves,
- controls, and accessories, and provides for pressurization of the drywell and the suppression chamber.
The drywell is thereby maintained at a
pressure ranging from 0.5 to 1.0 psig.
9.4.2.2.3 All Other Reactor Building Areas The HVAC subsystem is shown schematically on Figure 9.4-8.
The system has the following modes of operation:
Normal operation.
2.
'mergency operatio'n.
Normal 0 eration The supply ventilation air handling unit assembly consists of an air intake, prefilter, filter, heating
~ coil, cooling
- coil, dampers,
- controls, and supply fans.
Three 50-percent capacity vaneaxial fans are provided; two operate normally while one is in standby.
The prefilter and filter are of the extended surface di'sposable type.
The glycol heating coil preheats the supply air to the required discharge air temperature.
Glycol is supplied. to the heating coil from the plant glycol heating system (Section 9.4.11).
The cooling'oil maintains the required discharge air temperature.
Cooling water is supplied to the cooling coil from the service water system (Section 9.2.1)..
- 9. 4-23'
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Nine Mile Point Unit 2 FSAR 2.
3.
A recombiner mixes the drywell atmosphere and the suppression chamber atmosphere.
Prior to initiation of the recombiner, the drywell and the suppression chamber will be mixed uniformly due to natural convection and molecular diffusion.
Mixing will be further promoted by operation of the containment sprays.
The operator actuates the containment sprays within 30 minutes after the LOCA.
The criteria for the operation of containment sprays is specified in Section 6.2.1.1.
The recombiners will be started manually by the operator when the hydrogen or oxygen concentration exceeds a value of 4.5 volume percent.
An alarm is provided to aid the operator when d'rywell and suppression chamber monitors indicate a value of 4.5 volume percent of hydrogen or oxygen.
Two identical Category I recombiners are provided to limit oxygen or hydrogen concentration.
Operation of either recombiner will limit combustible gas concentration to a safe value.
- 1. 10
- 1. 11
- l. 13
- 1. 15 1.16 l
- 1. 17
- 1. 18
- 1. 19 1.20 1.21 1.22 1.23 1.24 6.
The components of the CGCS are protected from missiles and pipe whip to assure proper operation under accident conditions as required for safety-related systems.
The recombiners and monitors are located outside the primary containment.
The components ef the CGCS are designed as Category I and Safety Class 2.
1.25
- 1. 26.
1.27 1.28 7.
All components that are subjected to primary containment atmosphere will be capable of withstanding the humidity, temperature,
- pressure, and radiation conditions in the containment following a LOCA.
1.30 1.32 8.
The CGCS can be inspected or tested during normal plant conditions.
1.33 9.
The recombiners are located in the
- reactor bui lding.
1.34 Amendment 6.2-73 ch1217718f-14cz 09/05/84 155
l *f'
Nine Mile Point Unit 2 FSAR 5.
All controls for operating the CGCS (i.e., hydrogen recombiner system and monitoring system) are located in the main control room.
6.
A tabulation of the design and performance data for each system component is listed in Table 6.2-57.
1.. 13 1.14
- 1. 15 7.
8.
En'vironmental qualification information for safety-related equipment is given in Section 3.11.
. Electrical requirements for equipment associated with this system are in accordance with IEEE Class 1E standard.
The combustible gas control system is considered an
'extensi'on of the primary containment in post-LOCA conditions and consequently will be included within the boundary of the Type A test (Section 6.2.5).
The DBA hydrogen recombiner (HCS)
. system meets the criteria of Standard Review Plan 6.2.3 for closed loop systems as follows:
1.16 1.18
- 1. 19 1.20 1.27
- 1. 28 1.29 1.30 1.
Containment atmosphere does not directly communicate with the environment following a LOCAL 2.
Designed in accordance with Quality Group C
standards.
1.32 1.33 3.
Meets Category I design requirements.
4.
Is designed to primary containment pressure and tempe'rature design conditions as applicable.
S.
. Is-designed for protection against pipe
- whip, missiles, and jet forces.
6.
Is tested for leakage.
1.34 1.35 1.36 1.37
- 6. 2. 5. 2. 1 Atmospheric Mixing The function of post-LOCA mixing in the drywell and suppression chamber is performed by the primary containment spray
- system, recombiner
- system, and natural processes.
At approximately 30 min following the postulated
- accident, the redundant containment spray systems in the drywell and suppression chamber can be initiated to depressurize the containment.
The turbulence induced by the spray ensures a
.well mixed primary containment atmosphere.
In addition to the spray'ystem, the. blowdown of steam and water through the broken pipe creates a large degree of turbulence and promotes mixing of the entrained hydrogen and oxygen with Amendment 6.2-75
- 1. 43 1.44 I
- 1. 45
- 1. 46 1.47 1.49 ch1217718f-14da 09/05/84 155
Nine Mile Point Unit 2 FSAR to be greater than the bulk oxygen concentration.
The other two subcompartments are the control rod drive area in the drywell and the volume enclosed by the pedestal wall in the suppression chamber.
Due to the large open area between these two subcompartments and the bulk atmosphere, significant concentration gradients are unlikely.
6.2.5.2.2 Hydrogen Recombiner System The long-term control of hydrogen and oxygen is achieved by means of two identical 150-scfm thermal hydrogen recombiners',
Located in the reactor building and controlled from the main control room.
The recombiner system removes gas from the drywell or suppression
Flow from the suppression chamber atmosphere to the drywell through the vacuum breakers prevents the suppression chamber pressure from exceeding the drywell pressure by more than 0.25 psi.
Operation of any one recombiner will provide effective control over combustible gases within primary containment.
Figure 6.2-72a and b
shows the PAID of the recombiner system.
The manufacturer of the hydrogen recombiner is the Atomics International Division,, Energy Systems Group of Rockwell International.
1.12
- 1. 14
- 1. 15 1.16 1.18 1.19 1.22 1.23 1.25 1.26 1.27 1.34 1.35 1.36 1.37 The recombiner unit is skid mounted and is an integral package.
All pressure containing equipment including piping between components is considered an extension of the containment,
- and, therefoxe, is'esigned to ASME Section III, Safety Class 2 requirements.
The skid and the equipment mounted on it are designed to meet-Category I
requirements.
1.38 1.39 1.40 1.41 The recombiner unit consists of a blower, electric heater, reaction
- chamber, and water spray cooler.
The reaction chamber is capable of processing 150 scfm of gas containing up to either 2 1/2 volume percent of oxygen and unlimited excess hydrogen or 5 volume percent of hydrogen with excess oxygen.
Under these conditions, recombination efficiency is virtually 100 percent.
The recombiner is not designed to operate when hydrogen concentration exceeds 5 volume percent, with excess oxygen.
The recombination process takes place within the recombiner as a result of high temperature.
The resulting water vapor is then cooled along with other gases and returned to the suppression chamber.
- 1. 42
- 1. 43 1.44 1.45 1.46 1.47 1.48
- 1. 49
- 1. 50 Amendment 6.2-76 ch1217718f-14db 09/06/84 155
Nine Mile Point Unit; 2 FSAR The recombiner
- unit, which requires a
1 1/2-hr warmup period, is initiated manually from the control room prior to primary containment oxygen or hydrogen concentrations reach-ing 4.5 volume percent.
This occurs for the hydrogen con-centration, approximately 2.75 days after the design basis 1.52 1.53 1.55 Amendment
.6.2-76a ch1217718f-14db 09/06/84 155
Nine Mile Point Unit 2 FSAR 1.
Oxygen and hydrogen sources in a post-accident environment.
2.
Distribution of oxygen and hydrogen in the drywell and the suppression chamber.
3.
Primary containment pressure and temperature during the containment cooldown phase of the accident.
1.10 1.11 1.12 6.2 '.3.1 Sources of. Oxygen and Hydrogen Short-Term H dro en Generation In the period immediately after the I,OCA, hydrogen is generated by both radiolysis and metal-water reaction.
- However, the short-term 'contribution from radiolysis is insignificant compared. to that of the metal-water reaction.
The metal-water reaction of steam with the zirconium fuel cladding which produces hydrogen is:
Zr
+ 2H~O~ ZrO~
+
2H>
(6.2-14)
Based on, AVOCA calculational procedures and analysis of ECCS performance in conformance with 10CFR50.46 and Appendix K of
- 10CFR50, the extent of the chemical reaction is estimated to be 0. 14 percent of the fuel cladding material
~
The metal-water reaction generated hydrogen based on a core-wide penetration of 0.00023 inch results in a
metal-water reaction 'that is less than Jive times the calculated value of 0.14 percent (0.7 percent).
Therefore, 0.7 percent of the fuel cladding is assumed to react with water to produce hydrogen in accordance with Regulatory Guide 1.7.
The duration of this reaction is assumed to be 120 sec with a constant reaction rate.
The resulting hydrogen is assumed to be uniformly distributed in the drywell.
Figures 6.2-72D and 6.2-72E show hydrogen generation rates and integrated values as a function of time following the accident.
Short-Term Ox en Source 1.20 1.21 1.22
- 1. 24 1.25 1.26 1.27
- 1. 28 1.30 1.31 1.32
- 1. 34 1.35 1.36 1.37 1.38 1.39 1.40 1.42 The only source of air addition to primary containment, is the operation of relief valves inside the primary containment.
These relief valves are part of the breathing and service air systems, and. are normally isolated during reactor operation.
Due to high temperature following a AVOCA inside primary containment, a
portion of these systems (inside primary containment) becomes pressurized and relieves pressure by expelling about 126 standard cu ft of air into the primary containment.
1.43
- 1. 45
- 1. 47 1.48 1.50 Amendment 6.2-79 ch1217718f-14dc 09/06/84 155
Nine Mile Point Unit 2 FSAR The primary containment does not have any provision for storage of portable air packs for breathing.
The operating procedures would have appropriate controls for the use of portable ai r packs.
- 1. 51 1.52 Amendment 6.2-79a ch1217718f-14dc 09/06/84 155
Nine Mile Point Unit 2 FSAR THIS PAGE INTENTIONALLYBLANK Amendment
- 6. 2-.79b.
chl217718f-14dc 09/06/84 155
Nine Mile Point Unit 2 FSAR The automatic depressurization system (ADS) valves are nitrogen operated; therefore, operation of these valves will not result in addition of oxygen in the primary containment.
The short-term oxygen source has not been considered in the oxygen concentration evaluation, as it is very small.
1
~ 12
- 1. 13
- 1. 15
- 1. 16
~Eon -Term H dro en/Ox~en Generation
AVOCA.
Generation of hydrogen and oxygen due to radiolysis of core cooling water is an important factor in determining the long-term gas mixture composition within the primary containment.
A fission product distribution model as outlined in Reg'ulatory Guide 1.7 is used to calculate hydrogen/oxygen generati'on rates.
The incore radiolysis (due to core gammas) contributes hydrogen and oxygen to the drywell, and radiolysis due to fission products contributes hydrogen and 'xygen directly to the suppression chamber and the drywell atmospheres.
The division of hydrogen and oxygen between the suppression chamber and the drywell depends upon the fraction of water holdup on the drywell floor and water in the reactor'essel.
Hydrogen can also be formed by corrosion of metals and decomposition of organic materials in the primary containment.
The significant portion of this source is from the corrosion of zinc, which is included in the analysis'he temperature dependent hydrogen production rate is based on NUREG/CR -2812'"'.
The temperature-dependent hydrogen generation rate is shown in Table 6.2-59C for demineralized water.
The galvanized steel and zinc primer surface area exposed to sprays is shown in Table 6.2-59D.
The surface area used in the analysis is about 15 percent higher than the tabulated values.
The corrosion of aluminum in demineralized water is very small.
The Griess and Creek'est data suggest the hydrogen production rate to be between 4.76x10 to 3.23x10 Std.
cu ft of Hz per sq ft per hour.
Assuming that the corrosion in the Griess and Creek test is mainly due to 285'F and 212 F water temperature, the average rate is 4x10 'td.
cu ft of H~
per sq ft per hour.
Considering the aluminum surface area'irectly exposed to the spray environment and the above Hz generation
- rate, a
total of 125 SCF of hydrogen would be evolved within 20 days following,a I.OCA.
This being very small compared to other sources of hydrogen, Al corrosion and associated hydrogen production is ignored in the analysis.
- 1. 19 1.21 1.22 1.23
- 1. 24
- 1. 26 1.27 1.30 1.31 1.32 1.39
- 1. 40
- l. 41
- 1. 42
- 1. 43 1.44
=
- 1. 45 1.46
- 1. 47
- 1. 48
- 1. 49 1.50
- 1. 51.
1.52 1.53 1.54 1.55 Figures 6.2-72D through 6.2-72G
.show hydrogen and oxygen generation rates and integrated values.
The
. quantity of
- 1. 56
- 1. 57 Amendment 6.2-80 chl217718f-14dd 09/05/84 155
Nine Mile Point Unit 2 FSAR hydrogen initially contained within the reactor coolant system is negligible; hence, it is neglected.
6.2.5.3.2 Accident Description 2
~ 1 Following the postulated recirculation suction line double-ended rupture, the metal-water reaction begins in the core region and produces hydrogen immediately.
The reaction is assumed to last 2, min, during which 0.7 percent of the active zircaloy
. fuel: cladding reacts..
The radiolysis of coolant in the core region, water on the drywell floor, and suppression pool water begins immediately.
The hydrogen and oxygen thus generated evolve to the drywell and suppression chamber atmospheres.
2.2 2.7 2;8 Amendment 6'.2-80a ch1217718 f-14dd 09/05/84 155
Nine Mile Point Unit 2 FSAR THIS PAGE INTENTIONAjI,Y BIANK
.Amendment
.-6.2-80b ch1217718f-14dd 09/05/84 155
Nine Mile Point Unit 2 FSAR The combustible gases in the drywell and the suppression chamber would approach the flammability
- limit, if uncontrolled, after 4.75 days.
Prior to this, pressure and temperature within the primary containment are shown by analysis (Section 6.2.1) to have dropped to a level that will permit operation of the recombiner.
The recombiner system is manually activated when oxygen or hydrogen concentration reaches 4.5 percent.
The recombiner system takes suction from the primary containment atmosphere, recombines the hydrogen and oxygen to form water vapor, and returns the exhaust 'to the suppression chamber.
This results j.n a small pressure buildup in the suppression chamber that causes the opening.of the vacuum breaker valves between the drywell and suppression chamber.
As a
- result, the flow of the gas mixture from the suppression chamber to the drywell is established.
This arrangement. of recombiner suction and discharge promotes mixing of the two volumes in the primary containment.
6 '.5.3.3 Analysis 1.13 1.15 1.16 1.18 1.19 1.21 1.22 1.23 1.25 Based on the preceding hydrogen and oxygen generation sources and the accident description, the oxygen and hydrogen concentration in the drywell and suppression chamber is obtained as a function of time.
To calculate the redistribution of the hydrogen and oxygen between the drywell and suppression
- chamber, a two-region computer model of the primary containment system is used.
This model takes into consideration hydrogen and oxygen generation from the metal-water reaction and ~radiolysis.
The calculation determines the inventory, partial
- pressure, and mole fraction, of each atmospheric constituent in both regions as a function of time.
Tables 6.2-58, 59,
- 59C, and 6.2-59D present the parameters used in the analysis of the oxygen and hydrogen buildup within the primary containment.
Figures 6.2-72H and 6.2-721 present hydrogen and oxygen concentration transients in the primary containment, assuming only one recombiner is operating.
The recombiner is required to be functional approximately 2.75 days after the design basis accident.
1.26 1.28 1.29 1.30 1.32 1.33 1.35 1.36 1.37 1.39 1.40 1.41 Amendment 6.2-81 ch1217718f-14de 09/06/84 155
Nine Mile Point Unit 2 FSAR i
6.2.5.3 '
Failure Modes and Effects Analysis The failure modes and effects analysis (FMEA) for the CGCS is provided in the Nine Mile Point Unit 2 FSAR FMEA Report.
1.12
- 1. 13 6.2.5.4 Tests and Inspections 1.16 Each active component ol the CCCS is testable during normal reactor power operation.
This system will be tested periodically 'to assure that it 'ill" operate correctly whenever required.
Preoperational tests of the CGCS are conducted during the final stages of plant construction prior to initial startup.
These tests assure correct functioning of all controls, instrumentation, recombiners,
- piping, and valves.
System reference characteristics such as pressure differentials and
~ flow rates are'ocumented during the preoperational tests and will be used as base points for measurement.
in subsequent operational tests'uring normal operation, the recombiner system
- piping, valves, instrumentation, wiring, and other components can be inspected visually at any time, since they are outside the primary containment.
Further information may be found in Chapter 14.
1.17 1.19 1.20
'.22 1.23 1.24 1.25 1.26 1.27
- 1. 28
- 6. 2. 5. 5 Instrumentation Requirements 1.30 1.31 Safety-related instruments, and controls are provided for automatic and manual control of the hydrogen recombiners.
The controls and monitors described below are located in the main control room.
The control logic is shown on Figure 6.2-72K.
1.32 1.33 1.35 1'6 Instrumentation requirements for the primary containment purge. system and the SGTS portions ef the CGCS are described in Sections 9.4.2.5 and 6.5.1.5, respectively.
LOCA by means of the associated hydrogen recombiner LOCA override keylock switch.
1.42 1.44
- 1. 45 Amendment 6.2.-82 ch1217718f-14di 09/05/84 155
Nine Mile Point Unit 2 FSAR 6.2.7 References 1.
Models used in LOCTVS A Computer Code to Determine Pressure and Temperature
Response
of Vapor Suppression Containments Following a
Loss-of-Coolant
- Accident, Topical Report SWECO 8101, 1981.
- 1. 13 1.15
- 1. 16 1.17 2.
Maximum Flow Rate of a
Single Component Two-Phase Mixture, APED-4378, October 25, 1963.
1.18 3.
- Sharma, D.
F.
Technical Description Annulus Pressurization Load Adequacy Evaluation, NEDO-24548, January 1979.
4.
NUREG/CR-2812 (January 1984),
The Relative Importance of Temperature, pH 'and Boric Acid Concentration. on Rates of Hz Production from Galvanized Steel Corrosion.
1.19 1.20 1.22 1.23 5.
BNI-NUREG-24532 (Informal
- Report, May 1978),
Hydrogen Release Rates from Corrosion of Zinc and Aluminum.
1.25 Amendment
.6. 2-89 ch1217718f -14d j 09/05/84 155
Nine Mile Point Unit 2 FSAR TABIE 6. 2-59 PLANT PARAMETERS USED IN POST-DBA COMBUSTIBLE GAS CONCENTRATION ANALYSIS Reactor power Drywell free volume r
Suppression chamber free volume (at high pool water level)
Initial drywell pressure Initial drywell temperature Initial drywell relative humidity Initial suppression chamber pressure Initial suppression chambe r tempe rature Initial suppression chamber relative humidity Weight of-zircaloy in core (active fuel)
Zircaloy reaction with steam Duration of reaction Fraction of water. in drywell and reactor vessel Downcomer submergence at high pool water level 3, 467 MW 303,418 ft~
192,028 ft~
15.45 psia 135 F
40%
15.45 psia 90 F
100/
93,246 1bm 0.7%
120 sec 5.9%
11 feet 1.16 1.25 1.27 1.28
- 1. 30 1.32 1.34 1.36 1.37 1.39 1.40
- 1. 42
- 1. 43
- 1. 45
- 1. 46 1.48 1.50 1.52 J.
1.53 1.55 1.56 Vacuum breaker set, point Initial Oz concentration Recombiner-capacity 0.25 psid 1.58 150 scfm 2.4 4 volume percent
- 2. 2 pAmendment 1 of 2 ch1217718f-14df 09/05/84 112
Nine Nile Point Unit 2 FSAR TABLE 6.2-59 (Cont)
Recombination efficiency Temperature transient for primary containment (recirculation suction line DER)
~100%
Figure 6.2-8 2.6 2.9 2.10 2.11 2.12 Amendment 2,of, 2, chl217718f-14df 09/05/84 112
Nine Mile Point Unit 2 FSAR TABLE 6.2-59C CORROSION RATES Material Aluminum Corrosion Rate
~SCF~ft~-hrj 4.0 x.10
" (constant)
Applicable Temperature Up to 285~F 1.14 1:15 1
~ 16
- 1. 18 Zinc
-5113.25 0.6764 exp (460
+ T)
-23416.67 2.8245 x 10
exp 460
+ T) 1.20 119.12 F 5 T 5224.06 F
1.21 1.23 224.06 F 5 T 5 334.22 F
1.24 Amendment 1 of 1
ch1217718f-14dg 09g06y84 155
Nine Mile Point Unit 2 FSAR TABLE 6.2-59D AIUMINUM AND ZINC INVENTORY EXPOSED TO SPRAYS Material Aluminun Galvanized steel Zinc primer
,Surface Area gft~}
650 58,540',400 41, 500 6, 968 230 Amendment 1 of.l chl217718f-14dh 09/05/84 112
4 CJ tO C)
SuP
. Chas)kI Z
RadI OtP SI 5 Dyy~e l t R~d'>> n1y~I~
LLI CD lY CI x:
'2i Yl c.
CO<>OS IO>
C) pl 0
0 0
T I ME AFTER RCC IDENT
( SECONDS )
F I GURE 6 720 NTDIIODEII DENEAALIOH IIATES FOLLONIIID 051
Supp. Cha+
radio lysi ~
gyyu e l I P,hdiol s,is pl 0
p TINE AFTER ACCIDENT (SECONDS)
FIGURE 6-2-72E lNTK!hhMQ NTOAOOEH OENEhhTlON FOLL'ONlNO NA
4 VJ Ul CK C3 I
CK 4J J) ~~Q, I I P achi oipiS Wo+~ l O
~
g 0 pp. C 4Gwl P,adi Oiy S i5 TINE AFTER RCCI OENT
( SECONOS )
FlMJRE S.2-72F
~XYQEII OKÃEINTlON ltATES FOLLOMlHO OOA
ULl (0
K Cl I
CK CYbl g
hl C)
C3 4JI CT CY hl I
Q>y~ Q l I PAd i ol Sic 0
04 TIt1E AFTER ACCIOENT (SECDNOS)
FIGURE 6.2-72G TNTDhhTED OXYOKN ODINATION FOLLDMINO'eh
m/o Qf(.os Lal (JX C3tJ
~4 c>
5~~ af Supp-
~ceo
~ea
~sea
~eoe ~~
w~ee-2 3
+
G C
7
~
TINE RFTER RCCIOENT (SE&B%%)
bAQS FIGURE 6.2-72H OXTQEN COCOIThhTlOH FOLLOMlHO OOh
I UJ LJ UJ 0
,>r lt kQ(~M Irg n X
C3 I
CC QI 4J C3 UJ C3 lY C3 X:
)ty,l >n. r.
of Z.l 5 dog kg<,ornbin g S t-ac I-s Bn ye 8 oq +y v/0 Sv pP C 4 A'h k;. ~
S LIP P.
C4.'i rh Q ft$'>y~c l
t92866
~960
~866- ~~
~f499 3
G TINE AFTER ACCIDENT (88~0%)
MPs FIGURE 6.2-72I HYOIIOOEN CONCENTRATION FOLLOMTNO 05A
,)fV>SEX Nrne M>le Point Unzt 2
FSAR TABLE 6 2-56 CONTA.INMENT ISOLATION PROVISIONS FOR FLUID LINES Pene-tration System li.
~D Z-1A Main steam Line A GDC or Reg.
ESF Guide
~S stem Fluid Size
~in) 55 No Steam 26 FSAR Arrange-ment
~Fi ure(
~
6.2-70 Sh.
1 Location of valve Inside/
Outside Primary Contain-ment Inside outside length of Pipe Con-taanment to Outermost Isolation Valve 5'2w Type Test
((>
C C
Potential Bypass Leakage Path(>>
Yes Number 2MSS+HYV6A 2MSS+HYV7A Tel>e Ball Ball Oper-ator HYV HYV Hydraulic to open; spring to close N/A
~Pri mar secondary Actuator Mode Valve~ is)
Position Normal Shutdown Post-Accident closed Closed Open Isola-tion
~power Q) Signal Failure c+>
B,C,D, E,P,T, Rr RM closed Closure Time (5
e>
3 to 5 sec Note.
Power Source
( T>
N/A 8
drain line Z-1P Main steam Line 8 Main steam Line B
drain line Outside Outside 55 No Steam 26 6.2-70 Inside 5'-2" Sh.
1 Outside Outside C
C Yes 2MSS+HYV6B 2MSS+HYV7B Ball HYV HYV 2MSS*AOV938 Globe ACV 2MSS+AQV93A Globe AOV 2MSS4MOV208 Globe MOV Closed Closed Hydraulic N/A to open; spring to close closed Closed Open closed Closed Closed Closed Closed Closed Closed Closed Closed N/A FAI RM BrCro ~
FrErPr B,C,D, T,RM closed Closed RM 3 to 5
sec N/A N/A N/A N/A N/A Z-1C Z1D Z-2 Main steam Line C Main steam Line C
drain line Main steam Line D
drain line Main steam dram lane 55 No Steam 26
- 6. 2-70 Inside 5'-2" Sh 1
outside Outside 55 No Steam 26 6.2 70 Inside 5'-2" Sh.
1 Outside Outside 55 Steam 6
- 6. 2-70 Inside 1'-0" Sh.
2 outside C
C C
C C
C Yes Yes Yes 2MSS( AOV93C Globe AOV 2MSS+ HYV6D Ball 2MSS+HYV7D Ball HYV HYV 2MSS+AOV93D Globe Aov Hydraulic to open; spring to close Hydraulic to open; spring to close 2MSS*MOV111 2MSS+MOV112 MOV Elec.
MOV Elec 2MSS*HYV6C Ball HYV 2MSS*HYV7C Ball HYV N/A N/A Manual Manual Closed Closed Closed Closed N/A Open Closed Open Closed Closed Closed closed Closed Closed Closed B,C,D, R,RM RM 3 to 5 sec N/A 8
closed Closed B
C D
3 to N/A 8
E P
T 5 sec R,RM RM Closed Closed Closed FAI B,C ~ D Closed Closed Closed FAI Div II Div I Z
3 Spare TI APERTURE@
CARD 25' of 3>r
Docke", ~tt Contre
'E<~<
-'eument
-"'U'-KETFILp Nine Mile Point Unit 2
FSl TABLE 6. 2-56 (Cont)
Pene-tration System II.
~P-4A Feedwater line A to RPV GDC or Reg.
ESF Size Guide
~Sstem Fluid
~in 55 No Water 24 FSAR Arrange-ment
~iciuure( ( )
- 6. 2-7P Sh 3
Location of valve Inside/
Outside Primary Contain-ment Outside Inside Length of Pipe Con-tainment to Outermost Isolation Valve 2P 1<<
0 ~
0 It Type Test
())
C Potential Bypass Leakage Path(z)
W~)
Number 2FWS<<AOV23A 2FMS*V12A Tyi)e Swing Check Swing check Oper-ator AOV
~P Pneumatic Flow N/A Actuator Mode Normal (3)
Open Open (e)
Valve Position Post-Accident Closed Closed Shutdown Closed Closed Isola-
'ion Power))
Signal Farceur Reverse flow Closed Closed Closure Power Tzme Source (5
t))
(7)
The time,N/A it takes for one valve volume to pass through the valve Note/7 Z-4B Feedwater line A to RFV 55 No Water 24
- 6. 2-70 Inside OP-ott Sh.
3 Outside 2'1<<
N 2FWS'4V]2B Swing N/A Flow N/A m)
Check 2FMS*AOV23B Swing AOV Pneumatic N/A Check Open Open Closed Closed Closed Closed N/A N/A Reverse flow The time it takes for one valve volume to pass through the valve Z-5A RHS Pump A
suction from suppression pool 56 Yes Mater 24
- 6. 2-70 Outside 5'-6<<
Sh-4 No 2RHS+MOV1A Tricen-tric butter-fly MOV Elec Manual Open Closed Open FAI RM 45 Div I 13 Z-5B Z-5C RHS Pump B
suction from suppression pool RHS Pump C
suction from suppression pool 56 Yes Water 24 6.2-70 Outside 20P-g<<
Sh.
4 56 Yes Mater 24 6.2-70 Outside gt-gpt Sh.
Czg)
C No Lz'f )
2 RHS*MOV1 B 2RHS<<MOV1C Tricen-MCV tric butter-fly Tricen-MOV tric butter-fly Elec.
Elec.
Manual Open Closed Open FAI RM 45 Manual Open Closed Open FAI RM 45 Div II 13 Div II 13 Z-6A RHS test line Loop B to sup-pression pool 56 Yes Mater 18 6 2-70 Outside Sh 6
9 '
1 5/16<<C No (zy )
2RHS+MOV30B Tricen-tric butter-fly MOV Elec.
Manual Open Closed Closed FAI RM Div I 15 APERTURE CARD 2 of~
$ So9mumeu-oa
Pene-tration No.
Z-68 System
~dHS test line Loop A to sup-pression pool GDC or Reg.
ESF Guide
~s stem Fluid Size
~in 56 Yes Water 18 FSAR Arrange-ment
~Fi ure((
- 6. >-70 Sh.
6 Location of valve Inside/
Outside Primary Contain-ment Outside Length of Pipe Con-tainment to Outermost Isolation Valve 9 ~
3 dd Type Test
(()
Potential Bypass Leakage Path(z) c Number 2RHS+MOV30A T~oe Tricen-tric butter-fly Oper-a tor MOV P
d Elec.
~dd Manual Actuator Mode X4c2( I
IL@
Open Closed Closed Valve('I1 Position Normal Post-( ) >
shutdown Accident Nine Mile Point Unit 2 FS)
TABLE 6 256 (Cont)
Power Source
( z)
Div II Noti 15 Isola-tion Pouter l@ Sianal Failure (n)
FAI Closure T).me Z-7A RHS containment spray Loop A
to suppression pool 56 Yes Water
- 6. 2-70 Sh.
7 Outside 1st 3n 2RHS+MOV33A Globe MOV Elec.
Manual Closed Closed Open FAI G
23 Div I Z-7B RHS containment spray Loop B
to suppression pool 56 Yes Water 4
6 270 Sh.
7 Outside 4 ~
6 II C
No 2RHSnMOV33B Globe MOV Elec.
Manual Closed Closed Open FAI G
23 Div II 7/( 15 Z-BA RHS containment spray Loop A
to drywell 56 Yes Water 16 6 '-70 Sh.
8 Outside Outside 2 ~ Q II C
No 2RHs<<Mov25A Gate MOV Elec.
c<<)
Manual Closed Closed Open FAI RM 89 Djv I ZBB Z 9A RHS containment spray Loop B
to drywell RHS/LPCI LooP A to RPV 56 Yes Water 16 55 Yes Water 12
- 6. 'r-70 Sh.
8 6-:-70 Sh.
9 Outside Outside Outside Inside 2 ~ 3 n 7)
Qn C
C No~ g 2RHS+MQV24A Gate MQV 2RHS+AOV16A Check ACV Elec Process Nog ~)
2 RHS< MOV25 B Gate MOV Elec.
Manual Closed Closed Open FAI RM 89 Div II Div I Manual N/A Closed Closed Closed Closed Open Open FAI Closed RM 19 5
Reverse ~-o flow 15 15 Z
9B RHS/LPCI Loop B to RPV 55 Yes Water 12
- 6. 7-70 Outside 6'-6<<
Sh.
9 Inside C
No i
2RHS*MOV24B W 2 2RHS+AOV16B Check AQV Elec Process Manual N/A Closed Closed Closed Closed Open Open FAI Closed RM
- 19. 5 Reverse 9 g-o flow Div II N/A 1
15,11 Z-9C RHS/LPCI Loop C to RPV 55 Yes Water 12
- 6. 7-70 Outside 6)-6>>
Sh.
9 Inside C
C No 2RHS<<MQV24C g'g) 2RHS<<AOV16C Check AQV Elec.
Process Manual N/A Closed Closed closed Closed Open Open FAI Closed RM 19 5
Reverse S 9'O flow Div II N/A
)
15,1) )
TI APERTURE CARD 3 off' 509MQ962 -cS
+o P//o
()nt>))1 g <
REGULA,,
)- -on Do
" '-7'Fil,<
Nine Mile Point Unit 2
FS>
TABLE 6. 256 tcont)
Pene-tration No Z-10A System
~OHS shutdown return Loop A
to reactor re-circ Loop A
GDC or Reg.
ESF Size Guide System Fluid
~in 55 No water 12 FSAR Arrange-ment
~Fi ure<<>
6.2-70 Sh.
13 Location of valve Inside/
Outside Primary Contain-ment Outside Inside Length of Pipe Con-tainment to Outermost Isolation Valve 6 ~
pn T'I/pe Test
((>
C Potential Bypass Leakage Path(e>
Number TPfRe 2RHSnNOV40A Globe 2RHSnAOV39A Check Oper-ator MOV ACV Actuator
~P Elec.
Process Mode secon~dar Manual (g>
Valve Position Post-Accident Closed Open Closed Open Closed closed Normal Shutdown PowerCR Failure FAI Closed Isola-tion signal (t )
A,L,MD RM Reverse flow closure Time 25
@S 0 Power Source
( >)
Div I Div I Not(
RHS shutdown cooling return line inboard valve bypass line 55 No Water 2
- 6. 2-70 Inside Sh.
13 2RHSnNOV67A Globe MOV Elec.
Manual Closed Closed Closed FAI A ~ L,M~
RM Div I Z-10B RHS shutdown return Loop B
to reactor re-circ Loop B
55 No Water 12
- 6. 2-70 Sh.
1]
Outside Inside 6I pn C
No 2RHSnNOV40B Globe MOV Elec.
Manual (z't >
2RHSnAOV39B Check AOV Process N/A Closed Open Closed Open closed closed FAI Closed A,L,NP 25 RM Reverse&5-(3 flow Div I Div I RHS shutdown cooling return line inboard valve bypass line 55 No Water 2
- 6. 2-70 Sh.
13 Inside 2RHSnMOV67B Globe MOV Elec.
Manual closed closed Closed FAI A,L,M, 9
RN Div I Z-11 Z12 RHS shutdown supply from reactor recirc CSH suction from suppres-sion pool 55 No Water 20 6.2-70 Sh.
14 56 Yes water 20
- 6. 2-70 Sh.
5 Outside Inside Inside Outside 6 ~
pn 2 ~
2 II C
C C
(Zg )
/ e.s
~go)
MOV MOV HSnRV152 Relief N/A 2CSHnMOV118 Gate MOV 2RHsnMOV1 13 Gate 2RHsnMOV112 Gate Elec Elec.
Auto Elec.
Manual Manual N/A Manual Closed Open Closed Closed Closed Closed Closed-Closed Open Closed Open closed FAI FAI Closed FAI Div I Div II N/A A,LE iM, 27 RM A,L,M, 27 RN N/A N/A RN 16 Div III Z-13 CSH test return to suppression 56 Yes Water 12
- 6. 2-70 Sh.
15 Outside 5Q ~
Qn No 2CSHnMOV111 (z'f ')
Globe MOV Elec.
Manual Closed Closed Closed FAI 8, F,RM STD Div III HPCS min flow bypass Yes Water 4
Outside 45I 6n 2CSH*MOV105 Gate MCV Elec Manual Closed Closed Closed FAI RN STD Div III TI APRRTURI,'ARD 4 of 2(0
PLV>Sel Nine Mile Point Unit 2 Fs!
Pene-tration No System
~DDC or Reg.
ESF Guide
~s stem Fluid Size
~in FSAR Arrange-ment Fixture<<>
location of valve Inside/
Outside Primary Contain-ment Length of Pipe Con-tainment to Outermost Isolation Valve Type Test (I >
Potential Bypass Leakage Path<z>
Number Oper-ator Actuator Mode secon~dar
+l -'//r>
n n>J //
&Ayf2~~
WRy D'OC ocun>enf ET FQ@
(4)
Valve Position Normal Post-( >>
Shutdown Accident Isola-tion P~ove lO) Signal Failure C~
TABLE 6. 2-56 (Cont)
Power Source t>>
Mott Closure Time CS II>
Z-14 Z-15 CSH to RPV CSL suction from suppres-sion pool 55 Yes water 12 56 Yes Water 20
- 6. 2-70 Sh.
9
- 6. 2-70 Sh.
4 Inside Outside Outside 1 I 8II 2 CSL+ MOV1 1 2 Butter-fly 2CSH~AOV108 Check 2CSH~MOV107 Gate AOV MOV MOV Process Elec.
Elec.
Air Manual Manual Open Open Closed Closed Closed Closed Open Open Open Closed
,Reverse flow FAI RM FAI RM 27 90 N/A 1'3 Div III Div I 13,-
Z-16 Z-17 CSL to RPV ICS suction from suppres-sion pool 55 Yes water 12 56 Yes Water 6
- 6. 2-70 Sh.
10
- 6. 2->0 Sh 5
Inside Outside Outside 9 II C
fCa Air 2CSL+MOV104 Gate 2ICSI'MOV136 Gate MOV Elec.
Manual MOV Elec.
Manual 2CSL~AOV101 Check AOV Prccess Closed Closed Closed Closed Closed Closed Open Open Open Closed FAI FAI Reverse N/A flow RM 637 RM 30 N/A 11 13 Div I 120VDC Z-18 Z-19 ICS minimum flow to sup-pression pool ICS turbine exhaust to suppression pool 56 Yes Water 2
56 Yes Steam 12 6.2-0 Sh.
'1
- 6. 2 70 Sh.
12 Outside Outside 6 II2.75'o 2ICS+MOV143 Globe MOV Elec gz'l)
C No 2ICS~MOV122 Gate MOV Elec.
(z~)
Manual Manual Closed Closed Open FAI RM 5
Open Open Open FAI RM 61 120VDC 120VDC ICS turbine exhaust vacuum breaker 56 Yes Steam 1
1/2
- 6. 2-70 Outside S11.
12 2ICS4MOV164 Globe MOV Elec.
Manual Open Open Open FAI F~RM ~ H 10 f
Z-20 Spare No 3/4 TI APERTURE CARD 5 of+
Pene-tration No.
Z-21A System
~0team to ICS turbine and RHS heat exchangers Size
~in Yes Steam 10 GDC or
- Reg, ESF Guide
~sstem Fluid FSAR Arrange-ment
~F'.2-70 Sh.
16 Location of valve Inside/
Outside Primary Contain-ment Outside
- Inside, Length of Pipe Con-tainment to Outermost Isolation
- Valve 9N 1 II Type Test
( <)
C C
Potential Bypass Leakage Path(n)
Number Oper-ator 2ICS*MOV121 Gate NOV 2ICSsNOV128 Gate MOV pPrimary
~dd Elec.
Elec.
Manual Manual Actuator Mode Normal (s)
Open Open Nine Mile Point Unit 2 I TABLE 6. 2-56 (Cont)
Closure Time
<s e)
Power Source
=.
Not Div I 0 '//o Control P /<'md~of 13ocun)en 13ate- ~/ Y llpCKET EEOC"'osition Powers t n )
Post-Accident Shutdown Failure Closed closed Open Open
~ FAI FAI Isola-tion Signal
<4)
KERN K,RN
- 14. 5 14 5
ICS turline steam supply bypass to inboard isolation va lve Steam Inside 2ICS*MOV170 Globe.'OV Elec.
Manual Open Closed Open FAI K,RM 10.5 Z-218 Spare Z22 ICS to RPV 55 Yes Water 6
- 6. 2-70 Outside 4.25" Sh.
17 Inside No 2ICsd'AOV156 Check AOV Process (29) 2 ICS*AOV157 Check AOV Process Air Air Closed Closed Closed Closed Open Open Closed Closed Reverse 5
flow Reverse 5
flow 120VDC Z-23 WCS supply from RCS 8
RPV 55 No Water 8
- 6. 2-70 Inside 15" Water 8
Sh.
18 Outside t'CSg,i N
l 2WCS< MOV102 Globe MOV Elec.
2WCS< MOV112 Globe MOV ELec.
Manual Manual Open Open Open Open Closed Closed FAI FAI U, B,RM 14 U,Bd Jd 14 Div I Div I Z-24 Z-25 Z-26 Z-27 Spare RDS lines to RPV 53 Insert
.53 Withdrawal RDS lines to RPV 39 Insert 39 Withdrawal RDS lines to RPV 54 Insert 54 Withdrawal Yes Yes Yes Water 1
N/A 3/4 Water 1
N/A 3/4 Water 1
N/A 3/4 Outside 125'utside,125
~
Outside 125'utside 125'utside 125'utside125'/0
(~>)
u/0 C2~)
See Note 17 See Note 17 See No e 17 TI APgg~
CARD 6 of/
f617i5<PI Nine Mile Point Vnit 2 FSA Pene-tration No.
System
~tt'DC or Reg.
ESF size Guide
~S stem Fluid
~in)
FSAR Arrange-ment location of valve Inside/
Outside Primary Contain-ment Length of pipe Con-tarnment to Outermost Isolation Valve Type Test
<13 Potent1.al Bypass Leakage Path<<'>
Number Oper-ator'ctuator Mode Primary secon~dar KD Y/C3 Yo F2~
1e'7 337~P T't77@
(g)
Valve Normal (31 Shutdown Isola-Position tion post-PowerWIO) Signal Accident Failure TABLE 6 2-56 (Cont)
Closure Power Time Source Note Z-28 RDS lines to RPV 39 Insert 39 Withdrawal Yes Water 1
N/A 3/4 Outside Outside 125'25'ee Note 17 Z-29 SLCS to RPV 55 Yes Boron 1
1/2 solu-tion Inside Outside Outside No 2SLS'tv10 2 S Lse'OV5 A 2SLS+MOVSB Check Stop check globe Stop check globe N/A MOV MOV Process Elec.
Elec.
N/A Manual Manual Closed Closed 'losed Closed Closed Closed Closed C.BOA flu~
Ru/CN' y/aw 0/A t/1/A Z-30A Spare Z-30B Spare Z-31A Z-31P Z-31C Z-31D Z-31E Z-32 Z-33A TIP drive guide tube to RPV TIP drive guide tube to RPV TIP drive guide tube to RPV TIP drive guide tube to RPV
'IIP drive guide tube to RPV Nz purge to TIP index mechanism CCP supply to RCS Pump A
57 No Note 19 57 No Note 19 1 1/2 1 1/2
- 6. 2-70 Sh.
1'j
- 6. 2-70 Sh.
19 57 No Note 19 1 1/2
- 6. 2-70 Sh.
19 57 No Note 19 1 1/2
- 6. 2-70 S11 19
, 57 No Note 19 1 1/2 6.2-70 Sh.
19 No Nz 57 1 1/2 '
- 6. 2-70 Sh 20 56 No Water 4
Outside outside Outside Outside Outside Outside outside Outside Outside Outside Outside ea ches t77 K Inside Outside 1 '-3/4" 1
~ -3/4" 1 ~ -3/4w 1 e-3/4'e 1
~ 3/4u 1
~ 0 ee u.3'-o" 7 ~
0 ee C
C
"(3)
(3/)
No
( )
No
( 37)
No
( 31)
No (3 I) f 31)
No
( 3I)
N/A N/A N/A N/A N/A N/A N/A N/A 2CCPe'MOV94A 2CCPe'MOV Ball shear Ball Shear Ball Shear Ball Shear Ball Shear Check C
U-13'lobe Globe SCV N/A SOV N/A SOV N/A SCV N/A SOV N/A Simple check sc,v MOV MOV Elec.
N/A Elec.
N/A Elec.
N/A Elec.
N/A Elec.
N/A N/A Elec.
Elec.
Elec.
N/A Elec.
N/A Elec.
N/A Elec N/A Elec N/A N/A Manual Manual Closed Open Closed Open Closed Open Closed Open Closed Open oeCQ
~os Open Open Closed Open Closed Open closed Open Closed Open Closed Open Closed Ooen Open Closed Open Closed Open Closed Open Closed Open Closed Open Closed cLC3 c.D Closed Closed Closed Open Closed Open Closed Open Closed Open Closed Open Closed FAI FAI weP iten
+w B ~ F,RM B,F,RM N/A N/A N/A N/A N/A N/A 2.z 23 i/7' 120
'45'p
'e773C 1 20.
1s-& QC
~73C.
1 20-1 z-s QC V+C,
~ 120-lz-> VP 1773 ct 1 20-.
1~ VfhC N/A.
l2C3 VAC Div II Div I 18, 19 18,19 18 ~ 19 lse19 18 ~ 19 76
~7 7R 7 0
c., '<y l)o unf/b/
P-Y/
of/)
Oppugn/en6 Q IZL//5<(
one Mzle Point Unxt 2
FSA1 TABIE 6. 2-66 (Conti Pene-tration No.
Z-33B Z-34A
- Z-348
--Z-35 System D
CCP to Rcs Pump B
CCP return from, RCS Pump B r, Spare size
~in 56 No Water 4
56 No Water 56 No Water 4
GDC or Reg.
ESF Guide
~sstem Fluid FSAR Arrange-ment
~iciure<L4
- 6. 2-70 Sh.
20
- 6. 2-70 Sh.
21
- 6. 2-70 Sh.
21 Location of valve Inside/
Outside Primary Contain-ment Inside Cutside Inside Outside Inside Outside I.ength of Pipe Con-tainment to Outermost Isolation Valve Type Test CL4 7 ~
0 II C
C C
C C
C 7 ~
Off r
7 ~
0 II Potential Bypass Leakage Path>>>>
..No (3/)
~No (3/)
No (3/)
Number 2CCP>>MOV94B Globe 2CCP>>MO 4B Globe i(A 2CCP>>MOVE)
Globe 2CCP>>MO~/I~-Globe
/6 IB 2CCP>>MO 1
Globe 2CCP>>MO
- Globe
/58 Oper-ator MOV MOV MOV MOV MOV MOV Elec.
Elec.
Manual Manual Elec.
-.E 1 ec=,.-
Elec.
Elec.
Manual
. Manual Manual Manual Actuator Mode secon~dar
,ta)
Position Normal Shutdown Open Open Open Open Post-Accident closed closed Open OpencI Open Open Open Closed Open -,;;;Closed r.-.
Open Closed Open Closed PAI FAI FAI..
FAI=,=
'AI FAI BE F,RM BrF ARM B ~ F,RM B;F,RM BE F,RM BE F,RM Isola-tion Powe Io)Signal Failure Closure Time (s
>>)
Power Source C z1 Div II 23 Div I Div II Dim(I Div II Div I 2323-,
/'3'3 Notes
.Z36 Service air to drywell
~
56 No Air 6.2-70 Sh.
22 Outside Inside 7 II NQ ( 3/)
2SAS>>HCV161 Globe 2SAS>>HCV163 Globe Manual Manual Manual Manual N/A N/A closed Closed Closed Closed Closed Closed closed Closed LMCrLC LMC ~ LC N/A N/A Div I Div II Z37 Z-38A Breathing air to drywell RDS to recirc pump A seal 56 55 No Air No Wa ter 3/4
- 6. 2-70 Sh.
22
- 6. 2-70 Sh.
23 Outside Inside Inside Outside Outside
<5I On C
C No (3 /)
'9'i 2AAS>> HCV1 3 4 2AAS>> HCV1 3 6 2RCS>>V60A 2RCS>>V90A 2RCS>>V59A Globe Globe Check Check Check N/A N/A N/A Plow Flow Flow Manual Manual Manual Manual N/A N/A N/A N/A N/A Closed Closed Open Open Open Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed Closed N/A N/A LMC ~ LC LMC,LC Reverse flow Reverse flow Reverse flow N/A N/A Div I Div II N/A Z-38B RDS to recirc Pump A seal 55 No Water 3/4
- 6. 2-70 Sh.
23 Inside Outside outside
( 5
~
0 It No 2RCS>>V60B
'z9 2RCS*V90B 2RCS*V59B Check Check Check N/A N/A Flow Flow Flow N/A N/A N/A Open Open Open Closed Closed r
Closed Closed Closed Closed N/A N/A Reverse flow Reverse flow Reverse flow
////i, N/A Z-39 Drywell floor drain tank vent line 56 No Air 6.2-70 Sh.
24 Inside Outside C
C 2DFR>>MOV121 2DFR>>MOV120 Gate Gate MOV MCV
.Elec.
Elec.
Manual Manual Open Open Closed closed closed Closed FAI FA'I B,F,RM B,F,RM 28 28 Div II Div I Z-40 Equi pme nt drains from drywell 56 No Water 4
- 6. 2-70 Sh.
24 Inside outside 4
~
2 1/2>>
C C
yes
(~o) 2DER*MOV119 2DER>>MOV120 Gate Gate MOV MOV Elec.
Elec.
Manual Manual Open Open Closed Closed closed Closed FAI FAI B,F,RM B,E,RM
- 21. 3 21 3
Div II Div I 8 of 24
Pene-tration System N
~O-41 Reactor coolant recirc to sample cooler Size
~in 55 Water 3/4 GDC or Reg ESF Guide
~s stem Fluid FSAR Arrange-ment
~iciure< 1 I 6-2-70 Sh.
25 Location of valve Inside/
Outside Primary Contain-ment Inside Outside Length of Pipe Con-tainment to Outermost Isolation Valve (2i Pv TYPe Test C1>
C C
Potential Bypass Leakage Path<>>
No Number 2 RCS4 SOV1 0 4 2RCS*SOV105 T~e Globe Globe Oper-ator SOV SOV Actuator Mode Secon~dar Elec.
Elec.
Closed Closed Closed Closed Closed Closed
~eo
.e~ >eh Valv Position Normal Post-Shutdown Accident Nzne Mrle Pornt Unzt 2
PSALM TABLE 6 2-56 (Cont)
Isola-tion Powe Io3 Signal Fai lur
< + 4 Power Source Note!
Div II Div I Closed B,F,RM Closed B,F,RM Closure Time (5
60 Z-42A Eire protection for reactor recirc pump 56 No Water 2
6.2-70 ch 26 Inside Outside C
C (3l )
2FPW+SOV219 2FPW*SOV21 8 Gate SOV Gate SOV Elec.
Elec.
N/A N/A Closed closed Closed Closed Closed Closed Closed Closed B,F,RM B,F,RM Z-428 Eire protection water for reac-tor recirc pump
(
Z-43 Drywell floor drains Z-44A Capped spare Z-448 Capped spare 56 56
'Water 2
Water 6-2-70 Sh.
26
- 6. 2-70 Sh.
27 Inside Outside In'side Outside C
C C
C (3()
yes
~No (~c3 2F PW+ SOV2 2 1 2FPWOSOV220 2DFR+MOV140 2 DFR*NOV139 Gate Gate Gate Gate
Elec.
Elec.
El ec.
N/A N/A Manual Manual closed Closed Open Open closed Closed Closed Closed Closed Closed Closed Closed closed Closed
'FAI FAI ELF,RN B,F,RM B,F ~ RN B,F,RM 14 2
14-2 Z-44C capped spare Z-44D Capped spare Z-44E Service air to drywell 56 Air 6.2-70 Sh.
22 Outside Inside 5u C
C No (pi')
2SAS+HCV160 2SAS*HCV162 Globe Globe Manual Manual Manual Manual N/A N/A Closed Closed Closed/
Open Closed/
Open Closed closed Closed LNC,LC N/A Div'I Closed LMC,LC N/A.
Div II Z-45 Equipment drain tank (2DER-'XK1) vent to drywell Z-44F Breathing air to drywell 56 No Air 2
- 6. 2-70 Outside 5v Sh.
22 Inside 56 No Air 2
- 6. 2-70 Inside Sh.
27 Outside C
C (3 O) 2 DERI'OV1 3 0 2DER4NOV131 Globe Globe No (3 ()
2AAS~HCV135 Globe 2AAS*HCV137 Globe MOV MOV Elec.
Elec.
Manual Nanual Manual Manual N/A Manual Manual Closed Closed Open Open Closed/
Open Closed/
Open Closed Closed Closed Closed LMC,LC N/A Div I Div II Closed Closed FAI B F RN 9I ~
Div II FAI B,F,RN
's.s Div I Closed Closed LMC LC N/A Z-46A CCP supply to drywell space cooler 56 No Water 8
6-2 70 Inside 7'
Sh.
28 Outside C
C 2CCP*NOV273 2CCP~NOV265 Gate Gate MOV MOV Elec.
Elec.
Manual Manual Open Open Open Open Closed Closed FAI B ~ F, RN 40 Div II, FAI B,F,RN 40 Div I 9 of~
Tl APERTURE CARD 8~ho v~so ~Q~
Docket Control g 6/~F~ ~.
Dae Docnn)ant I YD'OCR'zz z)
<<Vf>~g Nine Mile point Unit 2 F,
Pene-tration No System
~d-468 Capped spare Z-46C Eire protection water for con-tainment hose reel standpipe Z-46D Capped spare GDC or Reg.
ESF Size Guide
~S stem Fluid
~in FSAR Arrange-ment
~F'ocation of valve Instde/
Outside Primary Contain-ment See Note 20 Length of Pipe - Con-tainment to Outermost Isolation Valve Type Test
())
Potential Bypass Leakage Path(z)
Y) (Si)
Number TtfZ)e Oper-ator Valve Z~3 position Actuator Mode Normal Primar
~Sd dt td Il
'd Fost-Poweroa)
Eai lure Power Source (7) closure Time I~so a-tion Sional TABLE 6 2-56 (Cont]
Not Z-47 CCP return from drywell space cooler 57 No Water 8
- 6. 2-70 Sh.
28 Inside Outside I
7 I 3II C
C No
~
2CCP~MOV122 Gate MOV Elec 2CCP<<MOV124 Gate MOV Elec.
Manual Manual Open Open Closed Open Open Closed FAI FAI B ~ FiRN 40 BFFFRM 40 Div II Div I Div I Z48 Purge exhaust from drywell Z-49 Purge inlet to drywell Z-50 Purge inlet to wetwell Z-51 Purge exhaust from wetwell 56 No Air 14
- 6. 2-70 Sh.
29 Inside outside Ai(/gi 56 No i'r 14 6.2-70 sh.
29 Inside outside q)e 56 No Qr 12
- 6. 2-70 sh.
29 Inside Outside Arr 12 6
2 70 Inside Sh-29 Outs).de C
(3( '1 C3 I)
"'(3>)
Butter-fly Butter-fly Butter-fly Butter-fly AOV AOV AOV 2 CPS'OV1 0 7 2CPS*AOV1 0 5 AOV ACV Butter-fly Butter-fly AOV 2" PS*AOV109 2CPSI'AOV111 Butter-fly Butter-fly AOV AOV 2CPS+AOV108 2 C PS') AOV1 1 0 2CPS4AOV106 2 CPS+ AOV1 0 4 Pneu-matic Pneu-matic Pneu-matic Pneu-matic Pneu-matic Pneu-matic Pneu-matic Pneu-matic Manual Manual Manual Manual Manual Manual Manual Manual Closed Closed Closed Closed closed Closed Closed Closed Closed Closed Closed closed Closed Closed Closed Closed Closed closed closed BE F,RM 5
Closed B,F,RM 5
Closed closed 8
F RM 5
closed closed Closed Closed Closed B,F,RM 5
13 F
RM 5
Closed Closed B F RM 5
Closed Closed B,F,RM 5
Div II Div II Div I Div II Div I Div II Div I Z-53A Instrument air to ADS valve accumulators Z-52A Capped spare Z-52B Capped spare 1
56 No ir 1 1/2
- 6. 2-70 Outside N
Sh.
30 Inside 1
~
0 If 1
~
0 II C
C P'e~
M(3O>
2IAS~SOV164 Globe SOV Eleec.
2IAS V448 Check N/A Process N/A Open Open Closed Open closed Closed Closed B,F,RM CS + S O Closed B,F,RM N/A Div I N/A 2S 10 ofM
'n
~KRONUR CARO
kef g g
y/
Contrui g D
d5vm dry
'LAI" RY o Document DOCKET FIDE Nine Nile Point Unit 2 FSAR TABLE 6-2-56 (Cont)
Pene-tration No.
Z-53B System
~d' Instrument air to ADS valve accumulators 56 No GDC or Reg.
ESF Guide
~s stem Fluid size
~in 1 1/2 FSAR Arrange-ment Ficiure
6.2-70 sh.
30 Location of valve Insider Outside Primary Contain-ment Outside Inside Length of Pipe Con-tainment to Outermost Isolation Valve 1
~ P If 1
~ Pv Type Test C1 4 C
C ye Potential Bypass Leakage Pathez4 Number 2IAS*SOV165 2IAS*V449 Globe Check Oper-ator N/A Sov Actuator Mode~dd Elec.
Process N/A N/A Va~lve u7 Closed Closed Open Closed Open Open Position Normal Post-Shutdown Accident Closed Closed BE F,RN B,F,RM Isola-
+ion PowerL<<3Siqnal Failure Closure Time es az N/A Div I N/A Power Source Notes Z-53C Instrument air to NSRV accumu-lator tank 56 No
~r 1 1/2
- 6. 2-70 sh.
30 Outside Inside 1
~
P td 1 I Pll C
C (3O )
2I AS*SOV1 6 6 2IAS~SOV184 Globe Globe Sov Sov Elec.
Process N/A N/A Open Closed Open Open Closed Closed Closed Closed B,F,RM 4$ o B,F ~ RM Div I Div II Z-54A Capped spare Z-55A Z-55B Hydrogen recom-biner 1A supply to wetwell Hydrogen recom-biner 1B supply to wetwell 56 56 Yes Air Yes Air
- 6. 2-70 Sh.
31
- 6. 2-70 Sh 31 Inside Outside Inside Outside C
C C
C No No(3 )
2HCS*MOV4A Globe MOV Elec.
2HCS~MOV1A Globe MOV Elec 2HCS4MOV48 Globe MDV Elec.
2HCSrNOV1B Globe MOV Elec Manual Manual Manual Manual Closed Closed Closed Closed Closed closed Closed Closed Open Open Open Open FAI FAI FAI FAI B,F ~ RN B,F,RM B,FrRM B,F,RM 18 5
18.5
- 18. 5
- 6. 2-70 Sh.
31
- 6. 2-70 Sh.
31 Inside Outside Inside Outside C
C C
C No(+,i 2HCS*MOV6A Globe NOV Elec.
2HCS*NOV3A Globe MOV Elec.
No 2HCS+NOV68 Globe MOV Elec.
2HCS~MOV3B Globe MOV Elec.
Manual Manual Manual Manual Closed Closed Closed Closed Closed Closed Closed Closed Open Open Open Open
'FAI FAI FAI FAI Br'M 8',RM B,F,RN B,F,RN
- 18. 5
- 18. 5
- 18. 5 18.
5 Div I Div I 22 Div II Div II 22 Z-57A Hyrdogen recom-biner 1A return from wetwell 56 Yes Air 6.2-70 Sh.
31 Inside Outside C
C No
~
2HCS*NOV5A Globe MOV Elec.
2HCSl NOV2A Globe MOV Elec.
Manual Manual Closed Closed Closed Closed Open Open FAI FAI BrFrRN 18 5
B,F RM 18 5
Div I Div I 22 Z-578 Hyrodgen recom-biner 1B return from wetwell 56 Yes Air 6.2-70 Sh.
31 Inside Outside C
C No (
~
2HCS+MOVSB Globe NOV Elec.
2HCS4MOV2B.
Glube MOV Elec.
Manual Manual Closed Closed Closed closed Open Open FAI BE F,RM 18.5 Drv II FAI B,F,RN 18.5 Div II 22 Z58 Containment purge to dry-well 56 No Air 2
- 6. 2-70 Inside Sh.
29 Outside C
C No (3 i) 2CPS+SOV122 Globe SCV Elec.
2CPSrSOV120 Globe SOV Elec.
Manual Manual Closed Closed Closed Closed Closed Closed Closed B,F,RN 5
Closed B,F,RN 5
Div II Div I TI APERTURE CARD 11 of
Rc viseg EGV A,pnv 7>ntun>enf
<.KP<7 FILp Nine Mile Point Unit 2 FSAR TABLE 6.2-56 (Cont)
Pene-tration No.
Z-59 System
~0ontainment purge to wet-we 11 GDC or Reg.
Guide 56 ESF
~Sstem Fluid No Air Size
~in FSAR Arrange-ment
~F>
ure<
6.2-70 Sh.
29 I.ocation of valve Insider Outside Primary Contain-ment Inside Outside I.ength of Pipe Con-tainment to Outermost Isolation Valve Type Test I>>
C C
Potential Bypass Leakage Path<a>
No (3/)
Number 2 C Ps a SOV1 2 1 2CPS*SOV119 Globe Globe Oper-ator SOV SCV Actuator Elec.
Elec.
Manual Manual Mode secon~dar Valve( I>
Position Closed Closed Closed Closed Post-Accident Closed closed Normal Shutdown Closed Closed B r F ~ RM B ~ F ~ RM Iso a-tion Power 3 Signal Failure
(+ >
Closure Time Power Source 'ote, Div II Div I Z-60A CMS from dry-well 56 No Air 3/4
- 6. 2-70 Sh.
32 Inside Outside C
2CMS*SOV6 lA 2CMS+SOV60A Globe SOV Elec.
Elec.
Open Closed Open Closed B, F, RM
<1. 5 Div I Z60B CMS from dry-well Z-60C CMS to dry-well Z-60D CMS to dry-we 11 Z-60E CMS from dry-well Z60F CMS from dry-well Z-60G CMS to drywell Z-60B CMS to drywell 56 56 56 56 56 56 Yes Air No Air Yes Air No Air Yes Air No Air Yes Air 3/4 3/4 3/4 3/4 3/4 3/4 3/4 6.2-70 Sh.
32
- 6. 2-70 Sh.
32
- 6. 2-70 Sh.
32
- 6. 2-70 Sh.
32
- 6. 2-70 Sh.
32 6.2-70 Sh.
32
- 6. 2-70 Sh.
32 Inside Cutside Inside Outside Inside Outside Inside Outside Inside Outside Inside Outside Inside Cutside C
C No No L,)
" C~')
No C3,q No C3 q NoC5q No C )
2CMS+SOV24A 2CMS>SOV24C 2CMS+SOV63A 2CMS+SOV62A 2CMS*SOV33A 2CMS+SOV32A 2CMS 4 SOV6 1 B 2CMS+SOV60B 2CMSASOV24B 2CMS>SOV24D 2CMS~SOV63B 2CMSWSOV62B 2CMS*SOV33B 2CMSASOV32B Globe Globe Globe Globe Globe Globe Globe SOV SOV SCV SCV SOV SOV SOV Elec Elec.
Elec.
Elec.
Elec.
Elec.
Elec.
Elec.
Elec-Elec.
Elec.
Elec.
Elec.
Elec.
Open Open Open Open Open Open Open Closed Closed Closed Closed closed Closed Closed Open Open Open Open Open Open Open Closed B,ARM
<1.5 Closed B ~ F ~ RM
<1.5 Closed B F RM
<1 5
Closed B ~ F,RM
<1. 5 Closed B
F Closed Bi F RM
<'1 5
Closed B,F,RM
<1.5 Div I Div I Div I
'iv II Div II Div II Div II Z-61A Capped spare 3/4 Z-61B CMS from wet-we 11 56 Yes Air 3/4
- 6. 2-70 Sh.
32 Inside Outside NO/3 ~q 2CMS4SOV26A GloLe SOV Elec.
2CMS>SOV26C Elec.
Open Closed Open Closed BiFiRM
<1.5 Div I Z-61C CMS to wetwell Z-61D Capped spare 56 Yes Air 3/4 3/4 6.2-70 Sh.
32 Inside Outside No
( 3 I) 2CMS>SOV34A Globe SOV Elec.
2CMS+ AOV35A Elec.
Open Closed Open Closed B,F,RM
<1 5
Div I Z-61E CMS from wet-well 56 Yes Air 3/4 6.2-70 Inside Sh.
32 Outside NQ (3
~
2CMS4 SOV26 B Globe SCV Elec.
2CMS*SOV26D Elec.
Open Closed OPen C] osed B
F RM
<1 5
D>.v Il
'12 of~
APERTURE CARD ggORQ5085Q-/0
Pene-tration No.
System t'-67 TWS backu P C Si sltE Z-61F CMS to wetwell GDC or Reg.
ESF Sire Guide
~S stem Fluid
~in 56 Yes Air 3/4
~Air'0 FSAR Arrange-ment
~inure<<>
6 2-70 Sh.
32 Location of valve Inside/
Outside Primary Contain-ment Inside Outside length of Pipe Con-tainment to Outermost Isolation valve Type Test (te Potential Bypass Leakage Pathcer Number 2CMSeSOV34B 2CMS*SOV35B Globe Oper-ator SOV Docket'o-Control g rf,'v~/~<+ <
Document ate
OlRY DChCEET FILE Valvet+
Position Actuator Mode Normal Eost-
~P
~Sd Sh td A
'd t
Open Open Closed Elec.
Elec.
Nine Mile Point Unit 2 FSAZ TABLE 6.2-56 (Cont)
Closure Power Time Source Notes Closed B,F,RM
<1.5 Div II Isola-tion P owerr~ S zg na 1 Pailure ci)
--68 Z-69 Z-70 Z-71 Z-72 Z-73 Capped spare Spare Capped spare Spare capped spare RHS relief valve dis-charge to suppression pool 10 14 56 No Water 6
- 6. 2-70 Outside 48'-6" Sh.
33 No N
2RHSeRV108 RV 2RHSeRV20C N/A N/A N/A N/A N/A N/A N/A None N/A N/A..
Z-74 Z-75 Z-76 Z-77 Z78 Z79 Z80 Z-81 Z82 Capped spare Capped spare Capped spare Capped spare Capped spare Capped spare Spent fuel pocl cooling Capped spare capped spare 56 1 1/2 1 1/2 1 1/2 No Water 1 1/2
- 6. 270 Sh.
40 1 1/2 Outside Inside No q
2SFCeV203 Globe Manual Manual 2SFCA V204 Manual Closed Closed Closed Closed Closed N/A N/A TI APERTURI CARD zS 13 of Pf
Docket g 3o-)i'toll f
p, 3
P4
'ORY Df)CKQ, RT FILR Nine Mile Point Unit 2
FSAR TABLE 6. 2-56 (Cont)
Pene-tration No System
~Ot'DC or Reg.
ESF Size Guide
~Sstem Fluid
~in FSAR Arrange-ment
~F Location of valve Inside/
Outside Primary Contain-ment Length of Pipe Con-tainment to Outermost Isolation Valve T5Ipe Test CI)
Potentral Bypass Leakage Path)>>
Number TIRe Oper-ator Prima~r secondarv Actuator Mode Valvel+>
Position Normal Post-
<s)
Shutdown Accident Isola-tion Powe)Lm) Signal Failure closure Power Time Source c())
No~es
"-83 Z-85 Z-86 Z-87 Z88A Z-88B Z-89A Capped spare Capped spare capped spare Capped spare RHS safety valve discharge to suppression pool RHR safety valve discharge to suppression pool LMS from dry-well 56 Yes Steam 12
- 6. 2-10 Sh.
14 Outside 116'"
56 No Air 3/4
- 6. 2-70 Inside Sh.
35 Outside 56 Yes Steam 12
- 6. 2-70 Outside 106'2 3/4" A
Sh.
'4 Nog1)
No
~
2LMS4SOV152 Globe SOV Elec.
2I.NS*SOV153 Elec.
.) 't See Note (Q3 za See Note Q3 Closed Closed Closed Closed B ~ F,RM
<1.5 Div II Div I Z-898 Capped spare 3/4 Z-89C LMS from wet-well 56 No Air 3/4
- 6. 2-70 Inside Sh.
35 Outside No q
2LMSI'SOV156 Globe scv Elec 2LMS*SOV157 Elec.
Closed Closed Closed Closed B,F,RM
<1.5 Div II Div I Z-89D Capped spare 3/4 Z-90 Z-91A Z-91B ICS vacuum breaker Instrument air to drywell Instrument air to drywell 56 56 N
N P 1 1/2 1
1/2 Yes Air 1
1/2
- 6. 2-70 Sh.
36
- 6. 2-70 Sh.
37 6.2-70 Sh.
37 Cutside Outside Outside Inside Outside Inside 13' I
Qll 1 I 011 1 I Qll 1 I-Qll C
C
"'(n )
j IIX (a o) 2ICS* NOV1 4 8 2ICS*NOV164 2 IAS*SOV 1 67 2IAS*SOV185 2IASI'SOV168 2 IAS*SOV1 8 0 Globe Globe Globe Globe Globe Globe NOV NOV SCV SCV SCV SOV Elec.
Elec.
Elec.
Elec.
Elec.
Elec.
Manual Manual Manual Manual Open Open Open Open Open Open Closed Closed Open Open Open Open Open Open closed closed Closed closed FAI
'AI Closed Closed Closed Closed FIRM None Br FIRN B,F,RN BE F,RM BE F,RM 10 10
<5.0
<5.0
<5.0
<5-0 Div II Div I Div I Div II'ivI Div II Z-91C Capped spare Z-91C Capped spare 1 1/2 1 1/2 ApERTURE CARD 14 ofW
i;,
<nbj ~
YZCt
~Ye~
~>O@y 'fa ago~
gC.v'I S lA3 Nine Mile Point Unit 2 FS TABLE 6. 2-56 (Cont)
Pene-tration No.
System
~0DC or Reg.
ESF Guide System Fluid Size
~in FSAR Arrange-ment Fixture<<r location of valve Inside/
Outside Frimary Contain-ment Length of Pipe Con-tainment to cutermost Isolation Valve Type Test
<<a Potential Bypass Leakage Path<a>
Number Oper-Actuator Mode T~-
t
~P Seconder Valve( l3 Position Normal Post-
<ai Shutdown Accident Iso a-Closure Power(t(F) Signal Time Failure
>r Power source cv>
Not Z-92 Sp~~
No Z-96 25~
Z-98A Z-98B RHR relief valve discharge to suppression pool RHR relief valve discharge to suppression pool 56 Yes Water 3
6.2-70 outside 207 '6>>
Sh.
38 56 Yes Hater 3
6.2 70 Outside 89'-8" Sh, 38 gz9) 2CSL>>RV123 2CSL+RV105 2RHS*RV61A 2RHS>>RV110 2 RHS>> RV13 9 2C SHN RV1 1 4 2CSH>>RV113 2RHS*RV61B 2RHS*RV61C 2RHS>>RV20B RV RV N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A None N/A N/A N/A None N/A N/A Z-99A Hydraulic unit from recirc flow control valve HYV 17A (drain line) 57 No Hy-3/4
- 6. l-70 Outside
<5'-0" draulic Sh, 39 No
/
2RCS>>SOV68A Globe SOV Auto Remote manual OPen Closed Closed Closed B
F RM
<1 5
N/A,
'2l Z99B Z-99C Hydraulic unit to recirc flow control valve HYV 17A (open line)
Hydraulic unit to recirc flow control valve HYV 17A (pilot line) 57 No Hy-3/4
- 6. 1-70 Outside
<5 ~ -0>>
draulic Sh.
39 57 No Hy-3/4 6.2-70 Outside
<5'-0' draulic Sh.
39 N/A No r
2RCS>>SOV67A Globe SOV Auto (3i)
N/A No r
~
2RCS>>SOV66A Globe SOV Auto Ql Remote manual Remote manual Open Closed Closed Closed B ~ F,RM
<1 5
N/A,2<
Open Closed Closed Closed B F RM
<1 5
N/A 26 APERTURE CARD Ic 15 of9
9/z//~
~~qg~ oN
)3ocket 4
,control 4 I 1)oeun1e>
, ~DOCKm'F'~
l(EG~~
Nine Mile Point Vnit 2 FsAR TABLE 6. 2-56 (Cont)
Z-99D Hydraulic unit to recirc flow control valve HYV 17A (closed line)
Pene-tration System N
. O~t'ize
~in 57 No Hy-3/4 draulic GDC or Reg.
ESF Guide
~S stem Fluid FSAR Arrange-ment FiciureC 11 6.2-79 Sh 39 location of valve Inside/
Outside Primary Contain-ment Outside Length of Pipe Con-tainment to Outermost Isolation Valve
<5I 0n Type Test CC>
N/A Potential Bypass Leakage P
Oper<<
Number T~e ator 2RCSc'SOV65A'- Globe--"-
SOV Actuator Auto Mode Secon~dar Remote manual Valve w Position Normal Cs)
Shutdown Post4t'so a-tion Powey)E() signal Failure Open Closed Closed Closed B ~ F, RM Closure Power Time Source:
s a1 C zl
+:Notes-
<1 5, N/A, r26 Z-100A Z-100B Hydraulic unit from recirc flow control valve HYV 178 (drain line)
Hydraulic unit to recirc flow control valve HYV 178 (open line) 57 57 No Hy-'/4 draulic No Hy-3/4 draulic
- 6. 2-7(
Sh.
3'.2-70 Sh.
39 Outside Outside Dw
<51-0s No 2RCSSSOV68B Globe
-SOV Auto N/A No 2RCSSSOV67B Globe SOV Auto
(~A Remote Open manual Closed Closed Closed Remote manual Open Closed Closed closed B ~ F~RM
<1 5
N/A
(
j26 A
BeFrRM
<1 5 '/A, j; +26 Z-100C Z-100D Hydraulic unit to recirc flow control valve HYV 178 (pilot) line)
Hydraulic unit to recirc flow control valve HYV 178 (closed line) 57 No Hy-3/4 draulic
- 6. 2-70 Sh.
39 Outside
<5'-0" 57 Hy 3/4
- 6. 2-70 Outside
<5'-09 draulic S)1 39 N/A No 2RCS*SOV66B Globe.- SOV A
("
uto No 2RCSSSOV65B Globe SOV Autc
('3))
Remote manual Remote manual Open Closed Closed Closed B,F,RM ';<1.5,N/A
'26 Open Closed Closed Closed B F RM
<1 5
N/A 26 All instrument lines from reactor vessel R. G.
1.11 No Air/
3/4 Water 6.2-70 Sh.
41 Outside As close as possible to containment EF check EFV N/A valves Auto N/A Open Open Open Open Excess N/A N/A :
27 flow All instrument lines penetra-ting primary containment R G.
1.11 Air/
3/4 Water 6.2-70 Sh.
41 Outside As close as possrble to containment
" (I()
EFV N/A Auto N/A Open Open Open Open Excess N/A flow N/A
, 27
~To be supplied in an amendment