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
ACCESSION NBR:8409250352 DOC ~ DATE: 84/09/21 NOTARIZED;,YES DOCKET FAOIL:50-410 Nine Mile Point Nuclear Stations Unit 2i Niagara Moha 05000410 AUTH, NAME AUTHOR AF F ILIATION MANGANgC ~ VS Niagara Mohawk Power
~ NAME RECIPIENT AFFILIATION Corp'ECIP SCHNENCERiA, L,icensing Branch 2 R
Forwards resPonses to SER OPenItems 120r107r181r182ep142b
<'UBJECT:
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 'SIZEe TITLE: Licensing 'Submittal:,PSAR/FSAR Amdts 8, Related orrespondence'OTES:PNL icy FSAR S L AhlDTS ONLY, 05000410 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL' NRR/DL/ADL 1 0 NRR LB2 BC 0 NRR LB2 LA 1 0 HAUGHEYgM 01 }
INTFRNAL: ADM/LFMB 1 0 ELO/HOS3 1 0 IE FILE 1 1 IE/DEPER/EPB 36 3 3' IE/DEPER/IR8 35 1 1 IE/DQAS IP/QAB21 1 NRR/DE/AEAB 1 0 NRR/DE/CEB 11 1 1.
NRR/DE/EHEB 1 1 NRR/DE/EQB 13 2 2' NRR/OE/GB 28 2 2 NRR/DE/MEB 18 1 NRR/OE/MTEB 17 1 1 NRR/OE/SAB, 1 NRR/DE/SGEB 25 1 1 NRR/DHFS/HFEB40 1 1 NRR/DHFS/LQB 1 NRR/DHFS/PSRB 1 1 32'RR/OL/SSPB
- 1. 0 NRR/DS I/AEB ?6 1 1 NRR/DSI/ASB 1 1 NRR/DS I/CPB 10 1 NRR/DS I/CSB 09 1 1 NRR/DS I/ICSB 16 1 lt NRR/DSI/METB 12 1 1 NRR/DSI/PSB, 19 1 NRR/OS I/RSB 23 N
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NIAGARA MOHAWK POWER CORPORATION/300 ERIE BOULEVARD WEST, 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, C. V. Man n Vice President Nuclear Engineering & Licensing NLR:ja Enclosure xc: Project File (2) 840925O352 84092i voa
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UNITEO STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Hatter of Niagara Mohawk Power Corporation ) Oocket No. 50-410 (Nine Mile Point Unit 2) )
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.
York and County of ~
Subscribed and sworn to before me, a.
a Notary Public in and
, this g/- day for the State of of m~
New 1984.
Notary Public in and for ncaa a, County, New York M y Coranission expires:
CHRISTINE AUSTIN
-:: Notarv Pyblic in the State of Net Yorft y Commission Exprres March SO, l~
ualiFed in Onondaga Co. No. 47876&g
WfTNA 3N15Pi%
AY AN lo 8 JS e9 ni o:tdo'I e".!o)l
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Nine Mile Point Unit 2 FSAR QUESTION F470.7 (12.4) 1. 12 Section 12.4 indicates that the dose 'ssessment is in 1. 13 progress and that doses and details of the man-rem evaluation will be provided in an amendment. The bases and l. 15 details of the dose assessment, as specified in Regulatory Guide 1.70, Revision 3, and Standard Review Plan 12.3-12.4 1.17 (NUREG-0800), should be supplied. Either provide this 1.18 information or a schedule for submitting the information.
RESPONSE 1.21 See revised Section 12.4. 1.22 j Amendment QEcR, F470. 7-1 chl2 177 fqr14x 09/05/84 155 I
~8gpg250352
P Nine Mile Point Unit 2 FSAR QUESTION F471.11 1.10 Subsection 12.4.2.1, Man-Rem Evaluation, states that the 1.11 Man-Rem evaluation is in progress, the details of which will 1. 12 be included as an amendment to the Unit 2 FSAR. 1. 14 Provide this information.
RESPONSE
See revi,sed Section 12.4.
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 1.12 components and equipment containing radioactive fluids, and 1.13 to a lesser extent from the presence of airborne radionuclides. Inplant radiation exposures during normal 1.14 operation, refueling, and anticipated operational 1. 15 occurrences are discussed in Section 12.4.2. Radiation 1. 17 exposures at onsite locations outside the plant are discussed in Section 12.4.3.
12.4.1 Design Criteria 1. 19 The criteria for doses to plant personnel during normal 1.20 operation and anticipated operational occurrences, including 1.21 refueling,. are based on the requirements discussed in 1.22 .
10CFR20. The design radiation lev'els during normal 1. 24 operation, refueling, and anticipated occupational 1. 25 occurrences are shown on Figures 1'2,3-34 through 12.3-66. 1. 26 Radiation exposures to operating personnel are within 1.27 10CFR20 limits. Radiation protection design features 1.28 (Section 12.3) and the health physics program (Section 12.5) assure that the occupational radiation exposures (ORE) to- 1.30 operating personnel during normal operation, refueling, and anticipated operational occurrences are as low as is 1.31 reasonably achievable (A?ARA).
12.4.2 Exposures Within the Plant 1. 40 12.4.2.1 Man-Rem Evaluation 1. 42 The occupational radiation dose assessment for .Unit 2 is l. 45 performed using the guidelines of Regulatory Guide 8.19'i'. -1. 46 The bases for the annual man-rem estimates are Unit 1 1.48 operating data are modified to account for 1.49 differences and improvements in Unit 2. The projected 1.50 radiation dose rates throughout the plant facilities are based on assumed radiation conditions after 5 yr of plant 1.51 operation and expected radiation dose rates. Operational 1.52 data from several BWRs'~'which show that the average annual man-rem per unit over several operating years is =
1.53 948 man-rem per year) are presented in Table 12.4-12. These 1. 54 data indicate that, in recent years, occupational. radiation-exposures have. been much larger than the radiation exposures 1.55 reported for operating BWR plants in the mid-1970's. The 1.56 primary reason for the increase in radiation exposure. has been the increase in manpower necessary to support the 1.57 expanding special maintenance activities.
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 1.58 radiation exposures by work functions for a all8.19. BWRs over 2.1 2.2 several years as suggested in Regulatory Guide The average values indicate that operating BWR plants have approximately 76 percent annual occupational exposure 2.3 attributed to routine maintenance (40 percent) and special 2.4 maintenance (36 percent). In recent years, plant 2.5 modifications attributed to feedwater sparger repairs, inspection, repair and replacement of recirculation piping, 2.6 TNI lessons-learned modifications, and increased snubber and 2.7 pipe hanger inspections have contributed to the growing 2.8 amount of occupational radiation exposures associated with 2.9 special maintenance work functions. Design features 2.10 described in, Sections 12. 1 and 12.3 for the Unit 2 BWR 5 plant should minimize the special maintenance work 2.11 experienced. at earlier-designed operating BWR plants.
Unit 2 design improvements that are expected to reduce the 2.12 occupational radiation exposures include the following: 2,13 Incorporation of flush connections, on the CRD scram 2. 15 discharge volume header permits condensate flushing 2.16 of piping to minimize corrosion product holdup in a 2.17 high personnel access area.
- 2. Use of filtered'ondensate water for CRD hydraulic 2.18 fluid and the reactor recirculation pump seal purge 2.19 provides a clean water source that should extend pump seal life.
- 3. Installation of permanent hoisting system and 2.20 access platforms for the recirculation pumps, main 2.21 steam isolation valves, and safety-relief valves minimizes maintenance time in the drywell. 2.22 An improved refueling platform makes fuel handling 2.23 activities more efficient, therefore less time is 2.24 spent on the platform.
A multistud tensioner reduces the amount of 2.25 man-hours necessary to handle the reactor vessel 2.26 head studs.-
- 6. A new handling tool and platform for the removal of 2.27 CRDs from beneath the reactor vessel reduces crew 2.28 size and time spent in the high radiation area.
- 7. Improved fuel design minimizes the buildup of 2.29 radiation levels near reactor coolant systems and 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 2.31 system eliminates the special maintenance that was 2.32 required on older BNR recirculation piping due to stress corrosion cracking.
- 9. Inservice inspection access is improved'y remote 2.33 equipment development and access doors for reactor 2.34 vessel and nozzle weld inspection.
- 10. The main steam isolation valves are ball valves 2.35 with greatly reduced maintenance requirements and 2.36 smaller leakage rates which reduce the amount of man-hours spent servicing and inspecting the 2.37 valves.
- 11. A decontamination platform is provided to wash the 2.38 walls of the reactor cavity pit and internals pool 2.39 to minimize contribution from this source.
- 12. Use of separate shielded cubicles for locating 2.40 redundant components and highly radioactive 2.41 components minimizes radiation exposures during maintenance activities.
- 13. Use of mechanical snubbers should reduce the 2.42 frequency of necessary inspection compared to 2.4 hydraulic-.operated snubbers.
- 14. Installation of a CRD flush tank removes highly 2.4 radioactive corrosion and fission products from CRD 2.4 internals prior to rebuilding.
The occupational radiation exposure for Unit 2 is determined 2. 47 for each of the Regulatory . Guide 8.19ork function 2.48 categories by identifying specific tasks within each of the 2.49 seven work function categories and determining the time and manpower requirements for those tasks. This information is 2.51 used with the expected dose rates in areas where work is performed to determine the radiation exposure from each 2.52 activity. Tables 12. 4-5 through 12. 4-11 provide estimates 2.53 of occupational exposures based- on the identification of 2.54 specific tasks within each of the seven work function 2.55 categories: routine operations and surveillance, nonroutine operations and surveillance, routine maintenance, radwaste 2.56 processing, refueling, 'nservice inspection, and special 2.57 maintenance. Table 12.4-4 summarizes the occupational dose 2.58 estimates for the seven work functions. A comparison 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 3.2 plants data for the period of 1974-1979 (before the TMI 3.3 accident). The higher occupational exposures for the period 3.4 of 1980-1982 are not expected at Unit 2 because plant 3.5 modifications that caused the increases have been incorporated into the original design of Unit 2. 3.6 12.4.2.2 Estimates of Inhalation Thyroid Doses 3.8 Inhalation doses during full-power operations will be 3.10 negligible in every .area except the reactor, turbine, and radwaste building areas. Potential airborne activities for 3. 13 these areas are given in Section 12.2.2. These 3.14 concentrations are based upon data given in NUREG-0016 and EPRI'-495.'he inhal'ation thyroid doses that. result are 3.16 given in Table 12.4-2.
Thyroid dose rates in Table 12.4-2 are calculated according 3.17 to:
D =
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Nine Mile Point Unit 2 FSAR References 1. 10
- 1. Regulatory Guide 8.19> Occupational Radiation Dose 1.12 Assessment in f.ight-Water Reactor Power Plants Design 1.13 Stage Man-Rem Estimates Revision 1, June 1979.
- 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 18 f- 14dk 09/05/84 155
Nine Mile Point Unit 2 TABLE 12. 4-4 ESTIMATED OCCUPATIONAl RADIATION DOSE BY WORK FUNCTIONS FOR UNIT 2 Annual Dose Percentage of 1.15 Function Man-Rem Yr Total Dose 1.16 Routine Operations and 56.0 lo. 4 Surveillance Non-routine Operations 32. 0 1.21 and Surveillance 1.22 Routine Maintenance 191.0 9b.2 1.24 Waste Processing .54.0 10. P. 1. 26-Refueling 2~.0 1. 28 In."ervice Inspection 107.0 Qo 5 1. 30 Special Maintenance $ 5.0 l2. 3 1.32 Total 5ZB.O 100.0 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 Avg. Dose Rate Exposure No. t Dose
~Ae Y mrem hr ~F'me er ol'orkers
~Fre cere ~mrem Yr
~ Operal.ions Surveillance l. 18 Reactor Building 1.3 1.5 2/shift 4.0 1. 19 Turbine Building 4.0 2.0 2/shi f t 18.0 1. 20 Chemistry Surveillance 0.5 5.5 10 Da i ly 10. 0 1.22 Secur i'ty Surveys 1.0 0.50 1/hr 4.0 1. 24 Instrumentation and Controls 0.1 6. 00 40 1/day 9.0 1. 26 Radiation Protection 11.0 19. 0 1/wk 11.0 1. 28 Surve i I lance 1. 29 Tote I 56.0 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 AY9. Dose Ri1 to Exposure Iio. o f Dose A~ct vit mrem hr T~ime hr Workers rretruenc ~mrem r Equipment Operat,ions 1. 18 RWCU System 1.0 GO.O . 2 1/yr 0. 10 l. 19 Condensate System ~ 2 2 8.0 2 1/day 13.0 1.20 RIIS System 0.2 2;0 2 1/month O.ill . 1.21 SFC System 1.0 6.0 ~ 2 1/yr 0 OI 1. 22 ECCS System 0.2 2.5 2 1/month 0:01 1. 23 SLS System 1.0 3.0 2 1/month 0. 10 1. 24 Instrument Cal ibration 1. 26 Instrumentation and 0.2 6.0 40 I/day 18.0 1. 27 Cont,rois 1.28 Radiation Honitors 1. 30 Linearity Checks 1.0 50. 0 1 2/yr 1.31 Calibration 1.0 225.0 2 I/I.S yr 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)
Avg. Dose Rate Exposure No. Dose 1. 14
~AC ivi Y ~@rem hr T~ime hr of'orkers
~rre eche jmmrem r 1:15 Operation of Liquid Radwaste 0.5 ,6. 0 1/shi ft 10.0 1. 17 System 1. 18 Operation of Solid Radwaste 0.5 6.0 1/shift 10. 0 1. 20 System 1.21 DAW Compacting 3.5 6.0 2/day 31.0 1. 23 RadwasI,e Shipments 1.0 6.0 1/wk 1.0 1. 25 DAW Shipments 5.0 16.0 1/mont,h ~
. 2.0 1.27 Tota I 54.0 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 Avg. Dose Rate Exposure hr No. o Fre uenc r Dose r
- l. 15
~Time I'orkers
~mr em 1. 16 Reactor Disassembly 12.0 75.0 10 1/1. 5 6.0 1. 18 Reactor Assembly 12.0 150.0 10 1/1'. 5 12.0 1. 20 Fuel Unload 2.0 200.0 )/1. 5 2.0 1. 22 Fuel Load 2.0 180.0 1/1. 5 2.0 1. 24 Fuel Preparation 2.0 100. 0 1/1. 5 1.0 1. 26
- 1. 28 Tota I 23.0 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 Avg. Dose Rate Exposure No. of Dose 1. 15
~ee mrem hr ~Time hr Horkers Fre uenc r ~ere r 1. 16 Reactor Bui Iding - 100.00 ~ 150 7 1/1. 5 70.0 1. 18 Priaa ry'ontainment, 1. 19 Reactor Building- 10.00 580 1/refue I ing 31.0 1.21 Secondary Containment, 1.22 Turbine and Hiscel- 10. 00 110 1/1. 5 6.0 1. 24 laneous Buildings 1.25 1:27 Tota I 107. 0 1. 29 Amendment 1 of 1 ch1217718 f-14cu 09/06/84 155
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~ I e Nine Hile Point Unit 2 FSAR TABLE 12.4-12 OPERATIONAL HAN-REH PER YEAR fOR SELECTED BWR PLANTS
~PI an ~174 I 997 I 991 j ~17 7 ~178 ~17 1 9¹ ~18 1 ~182 Dresden 1,2,3 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 1,000 375 157 531 1,004 993 1.20 Nine Hile Point 824 681 428 1,383 314 1,497 591 1,592 1,264 1.22 Peach Bottom 2,3 228 840 2,036 I 317
~ 1,388 2,302 2,506 1,977 1.24 Quad Cities 1,2 482 1,618 1, 651 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 984 I, 140 1, 078 1, 614 1, 279 467 I 733
~ 917 865 1. 34 Brunswick 1,2 1,004 2,602I, 3,870 2,638 3 '92 1.36 Brown ferry 1,2,3 I, 792 1,66 I I 1,825 2,380 2,220 1.38 f i tzpatrick 1,080 909 859 2,040 1,425 1, 190 1. 40 Avg mrem/un i t 594 878 784 974 686 872 1,419 1,183 1,140 1.42 Overall avg. of 948 mrem/year-unit 1.44 i Rel'erence: NUREG-0713, Volume 4 1.46 Amendment 1 of 1 f
c h1 217718 -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 1.17 and Surveillance 12.3 13 ' 7.6 7' 9.1 10.0 1.18 Routine Maintenance 43.2 39.3 42.8 42.2 33.7 40.2 1.20 Waste Processing 5. 8. 4.3 3.1 11.0 6.2 6.1 1 22
~
Refueling 2.0 4.4 5.2 2.5 2.7 3.4 1.24 Inservice Inspection 2.6 7.3 3.3 3.7 4.3 4.2 1.26 Special Maintenance 34. 1 31 2
~ 38.1 33.1 44.0 36.1 1.28 1.30
References:
1.32 UREG-0594, "Occupational Radiation Exposure at Commerical l. 34 Nuclear Power Reactors, 1978," November, 1979. 1.35 NUREG-071/, Volume 1, "Occupational Radiation Exposure at 1.37 Commercial Nuclear Power Reactors; 1979," 1.38 March, 1981. 1.39 NUREG-0713, Volume 2, "Occupational Radiation Exposures "
of 1 41
~
Commer al Nuclear Power Reactors, 1980, 1.42 December, 1981 ~
1.43 NUREG-0713, Volume 3, "Occupational Radiation Exposure at 1.45 Commercial Nuclear Power Reactors, 1981," 1.46 November, 1982. 1.47 NUREG-0713, Volume 4, "Occupational Radiation Exposure at 1. 49 Commercial Nuclear Power Reactors, 1982," 1.50 December, 1983. 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 inc t udes the fol lowing features:
- a. 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.
- 2. 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).
Preparation of systems list, identifying specific methods used to test systems, the system involved, and frequency of testing.
- 5. 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 + wl Leak These mea ures a~@'pgImabg Qr mmes 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.
~~
~4 4
- 1.10-126
Nine Mile Point Unit'. 2 FSAR
- 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 the neutron it is necessary to lower 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 VOLUME 255 MARGIN Temp-Mat 122.25 ft. 3 153 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.
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- 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.
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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 12 reactor building leakage xate is within design limits.
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 12 intex.locks will be tested.
- 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 can maintain reactor building pressure equal to or below -0.25 wg.
- 5. The reactor building leakage rate is not greater than 3,160 cfm.
- 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 Systems lines may be excluded from con-leakage paths.
sideration as potential bypass paths for reasons such as:
the lines terminate in the secondary containment, an air water sealing system is provided to process or eliminate
'r 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
g<<ISED Nine Mile Point Unit. 2 FSAR 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 17 of~Z5
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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Rgb'~ P 1M'1C 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 nominal valve diameter closing rate is 12 for gate valves and 4 in/min of in/min o f 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 position upon failure of valve power remain in the as-is (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 addition, RHR Loops A and B have several discharge paths (LPCI, drywell 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'of lines are not part of the RCPB. The classification 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 i so 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 wi ll 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 assemblyat not leak at a rate greater than 10 'tm est'ng of the shear valves '
cc/sec 125 psig.
not Further leak 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 indicative pressure by a nitrogen purge. The purge makeup is 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 containment.
the primary
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.
s 22 of Q
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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:
Do not communicate with either the reactor coolant system or the containment atmospnere.
Are protected against m'ssiles and p'pe w¹p.
C. 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 1.17 line safety valves. 1.18 2RHS*RV56B - RHR heat exchanger shell side 1.22 relief valve. 1.23 2RHS*MOV26B and 2RHS*MOV27B RHR heat. 1.27 exchanger vent line isolation valves 1.28 2RHS*V117 and 2HS*V118 - RCIC vacuum breaker 1.32
, line. ~ 1.33 The valve is open only during steam condensing mode. 1.37 Valve position is indicated in the main control room to 1.38 provide the operator confirmation of valve status.
'Normally closed. Opened only when testing wetwell to 1. 42 drywell vacuum breakers.
'~~'Penetrations Z-99A,B,C,D, and Z-100A,B,C,D contain lines 1.45 for the hydraulic control of the reactor recirculation flow control valve. These lines contain- hydraulic fluid 1.47 used to position the reactor recirculation flow control valve.
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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7 Z> Egg Pl/(FAN 7/>Pp 5j Sfe&g Nine Mile Pbint Unit 2 FSAR QUESTION F480.29 (6.2.4) 1.12 Describe the provisions to insure that debris will not become entrained in the purge valves and prevent their closure. Guidance is provided below which, would represent an acceptable debris screen design:
if followed, a) The debris screen should be seismic Category I and 1.17 installed .typically about one pipe diameter away 1.18 from the inner side of the inboard isolation valve. 1.19 b) The piping between the debris screen and the valve 1.20 should also be seismic Category I design. 1.21 c) The debris screen should be designed to withstand 1. 22 the COCA diTferential pressure. 1. 23 d) The debris screen openings should be about 2 inches 1.24 by 1 3/16 inches. 1.25 A suggested debris screen design is enclosed as Figure 1. 1.27 RESPONSE 1.29 Debris screens, in general accordance with the design 1.30 referenced above as Figure 1, are provided. 1.31 See revised Sections 6.2.4.3.2 and 9.2.4. 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 1. 12 tank.
i normal plant conditions s from the condensate water storage 1.13 The elevations of the ECCS pump suction centerlines and the 1.14 suppression pool minimum drawdown level are 195 '0" and 1.15 197'-8", respectively. 1.16 Influent and Effluent Lines from Dx well and Su ression 1.18 Chamber Free Volume Primar Cont:ainment: Pur e'Lines The dxywell and 1.20 suppression chamber purge lines have isolation 1.21 capabj.lities commensurate with the impoxtance to 1.22 safety of iqolating these lines. Each line has two 1.2 normally ~osed/fai,l closed valves - one located inside (nitrogen operated) and one located outside 1.25 (air operated) the primary containment. The 1.26 inboard end of each 12-in and 14-in valve located inside the primary containment is provided with.a 1.27 QA Category I debris screen to prevent entrainment of foreign matter in the valve seat. The isolation 1.2 valves are interlocked to preclude opening of the 1.30 valves while a primary containment isolation signal 1.31 exists (Table 6.2-56).
- 2. Primar Containment Atmos here Monitorin S stem 1.32 Sam lin Lines The primary containment atmosphere 1.33 monitoring system consists
. of radiation and 1.34 hydrogen/oxygen moni toring lines. Each line, 1.35 suction and discharge, penetrates the primary containment and continuously monitors the radiation level and hydrogen/oxygen concentration during 1.36 normal operation. These lines are equipped with 1.37 two solenoid-operated isolation valves, one inside the primary containment and the other outside, 1.38 located as close as possible to the primary 1.39 containment. The hydrogen/oxygen monitoring lines 1.40 are also used to continuously monitor the pximary containment air during the post-LOCA period. Each 1.42 isolation valve receives isolation signals. The 1.43 isolation valves for hydrogen/oxygen monitoring lines are provided with individual keylock switches to override the isolation signal and initiate- 1.44 system operation, during the post-LOCA period.
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 1. 13 each floor level. The emergency recirculation system 1.14 ensures mixing throughout the reactor building atmosphere, 1.15 including the spent fuel pool area.
The intake duct connection for the SOTS (Section 6.5.1) is 1.23 taken at the discharge side of the emergency recirculation unit cooler to maintain the reactor building at a negative 1.24 pressure.
Unit space coolers with sufficient capacity to satisfy the 1.25 cooling requirements of the emergency safeguard equipment 1.26 provide cooling to handle the heat gain load of the 1.27 respective safeguard equipment. Cooling for general areas 1.28 is provided by unit ~ace coolers.
HVAC equipment and components that operate following a TOCA 1.30 are designed to Category I and Safety Class 2 and 3 criteria. Equipment motors and controls in the safety- 1.32 related portion of the system are supplied from their 1.33 respective independent emergency power sources and have 1.34 sufficient redundancy to satisfy the single-failure criterion. 1 ~ 35 9.4.2.3 Safety Evaluation 1.37 The safety features of the reactor building HVAC system are 1.39 as follows:
All safety-related components are designed to 1.42 Safety Class 2 and 3 criteria and Category I 1.43 requirements. Safety-related components are 1.44 located so that failure of a portion of other 1.45 nonessential systems does not prevent operation of any safety-related system. 1. 46
- 3. Redundant isolation valves in each line penetrating 1.51 the primary containment are in accordance with ASME 1.52 Section III. The piping between the isolation 1.53 valves is Safety Class 2 and both the valves and 1.54 piping are designed to Category I. All other 1.55 system piping is seismically supported. The 1.56 inboard end of each 12-in and 14-in CPS isolation valve located inside the primary containment is 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 f
ch12 177 18 -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 are performing properly. Flow and pressure if components 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 inerting system.
'y Oxygen control within primary containment during normal plant operation is achieved means of the nitrogen 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 system (cps) A4D ~
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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. Provi e a means o maintain~cQ the primary containment at positive pressure ~'uring normal operation so that any leakage can be monitored.
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.3. 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 di schargeh 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. A recombiner mixes the drywell atmosphere and the 1. 10 suppression chamber atmosphere. Prior to 1. 11 initiation of the recombiner, the drywell and the suppression chamber will be mixed uniformly due to natural convection and molecular diffusion. Mixing l. 13 will be further promoted by operation of the containment sprays. The operator actuates the 1. 15 containment sprays within 30 minutes after the LOCA. The criteria for the operation of 1.16 l containment sprays is specified in Section 6.2.1.1. 1. 17
- 3. The recombiners will be started manually by the 1. 18 operator when the hydrogen or oxygen concentration 1. 19 exceeds a value of 4.5 volume percent. An alarm is 1.20 provided to aid the operator when d'rywell and suppression chamber monitors indicate a value of 4.5 volume percent of hydrogen or oxygen. 1.21 Two identical Category I recombiners are provided 1.22 to limit oxygen or hydrogen concentration.
Operation of either recombiner will limit 1.23 combustible gas concentration to a safe value. 1.24 The components of the CGCS are protected from 1.25 missiles and pipe whip to assure proper operation under accident conditions as required for safety- 1. 26.
related systems. The recombiners and monitors are 1.27 located outside the primary containment.
- 6. The components ef the CGCS are designed as 1.28 Category I and Safety Class 2.
- 7. All components that are subjected to primary 1.30 containment atmosphere will be capable of withstanding the humidity, temperature, pressure, 1.32 and radiation conditions in the containment following a LOCA.
- 8. The CGCS can be inspected or tested during normal 1.33 plant conditions.
- 9. The recombiners are located in the -
reactor 1.34 bui lding.
Amendment 6.2-73 ch1217718f-14cz 09/05/84 155
l *f' Nine Mile Point Unit 2 FSAR
- 6. A tabulation of the design and performance data for 1. 15 each system component is listed in Table 6.2-57.
- 7. En'vironmental qualification information for safety- 1.16 related equipment is given in Section 3.11. 1.18
- 8. . Electrical requirements for equipment associated 1. 19 with this system are in accordance with IEEE Class 1E standard. 1.20 The combustible gas control system is considered an 1.27
'extensi'on of the primary containment in post-LOCA conditions and consequently will be included within the boundary of the 1. 28 Type A test (Section 6.2.5). The DBA hydrogen recombiner 1.29 (HCS) . system meets the criteria of Standard Review Plan 6.2.3 for closed loop systems as follows: 1.30
- 1. Containment atmosphere does not directly 1.32 communicate with the environment following a LOCAL
- 2. Designed in accordance with Quality Group C 1.33 standards.
- 3. Meets Category I design requirements. 1.34
- 4. Is designed to primary containment pressure and 1.35 tempe'rature design conditions as applicable.
S. . Is- designed for protection against pipe whip, 1.36 missiles, and jet forces.
- 6. Is tested for leakage. 1.37
- 6. 2. 5. 2. 1 Atmospheric Mixing The function of post-LOCA mixing in the drywell and 1. 43 suppression chamber is performed by the primary containment spray system, recombiner system, and natural processes. At 1.44 approximately 30 min following the postulated accident, the I redundant containment spray systems in the drywell and 1. 45 suppression chamber can be initiated to depressurize the containment. The turbulence induced by the spray ensures a 1. 46
.well mixed primary containment atmosphere. In addition to 1.47 the spray'ystem, the. blowdown of steam and water through the broken pipe creates a large degree of turbulence and 1.49 promotes mixing of the entrained hydrogen and oxygen with Amendment 6.2-75 ch1217718f-14da 09/05/84 155
Nine Mile Point Unit 2 FSAR to be greater than the bulk oxygen concentration. The other 1.12 two subcompartments are the control rod drive area in the drywell and the volume enclosed by the pedestal wall in the 1. 14 suppression chamber. Due to the large open area between 1. 15 these two subcompartments and the bulk atmosphere, significant concentration gradients are unlikely. 1.16 6.2.5.2.2 Hydrogen Recombiner System 1.18 The long-term control of hydrogen and oxygen is achieved by 1.19 means of two identical 150-scfm thermal hydrogen recombiners', Located in the reactor building and controlled 1.22 from the main control room. The recombiner system removes 1.23 gas from the drywell or suppression chamber, recombines the hydrogen with oxygen, and returns the gas mixture along with 1.25 the condensate to the suppression chamber. Flow from the 1.26 suppression chamber atmosphere to the drywell through the vacuum breakers prevents the suppression chamber pressure 1.27 from exceeding the drywell pressure by more than 0.25 psi.
Operation of any one recombiner will provide effective 1.34 control over combustible gases within primary containment.
Figure 6.2-72a and b shows the PAID of the recombiner 1.35 system. The manufacturer of the hydrogen recombiner is the 1.36 Atomics International Division,, Energy Systems Group of Rockwell International. 1.37 The recombiner unit is skid mounted and is an integral 1.38 package. All pressure containing equipment including piping 1.39 between components is considered an extension of the containment, and, therefoxe, is'esigned to ASME 1.40 Section III, Safety Class 2 requirements. The skid and the 1.41 equipment mounted on requirements.
it are designed to meet- Category I The recombiner unit consists of a blower, electric heater, 1. 42 reaction chamber, and water spray cooler. The reaction 1. 43 chamber is capable of processing 150 scfm of gas containing 1.44 up to either 2 1/2 volume percent of oxygen and unlimited 1.45 excess hydrogen or 5 volume percent of hydrogen with excess oxygen. Under these conditions, recombination efficiency is 1.46 virtually 100 percent. The recombiner is not designed to 1.47 operate when hydrogen concentration exceeds 5 volume percent, 1.48 with excess oxygen.
The recombination process takes place within the recombiner 1. 49 as a result of high temperature. The resulting water vapor 1. 50 is then cooled along with other gases and returned to the suppression chamber.
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 1.52 period, is initiated manually from the control room prior to primary containment oxygen or hydrogen concentrations reach- 1.53 ing 4.5 volume percent. This occurs for the hydrogen con- 1.55 centration, approximately 2.75 days after the design basis Amendment .6.2-76a ch1217718f-14db 09/06/84 155
Nine Mile Point Unit 2 FSAR
- 3. Primary containment pressure and temperature during 1.12 the containment cooldown phase of the accident.
6.2 '.3.1 Sources of. Oxygen and Hydrogen 1.20 Short-Term H dro en Generation 1.21 In the period immediately after the I,OCA, hydrogen is 1.22 generated by both radiolysis and metal-water reaction.
However, the short-term 'contribution from radiolysis is 1. 24 insignificant compared. to that of the metal-water reaction. 1.25 The metal-water reaction of steam with the zirconium fuel 1.26 cladding which produces hydrogen is: 1.27 Zr + 2H~O ~ ZrO~ + 2H> (6.2-14) 1. 28 Based on, AVOCA calculational procedures and analysis of ECCS 1.30 performance in conformance with 10CFR50.46 and Appendix K of 1.31 10CFR50, the extent of the chemical reaction is estimated to be 0. 14 percent of the fuel cladding material ~ The metal- 1.32 water reaction generated hydrogen based on a core-wide penetration of 0.00023 inch results in a metal-water 1. 34 reaction 'that is less than Jive times the calculated value of 0.14 percent (0.7 percent). Therefore, 0.7 percent of 1.35 the fuel cladding is assumed to react with water to produce 1.36 hydrogen in accordance with Regulatory Guide 1.7. The 1.37 duration of this reaction is assumed to be 120 sec with a constant reaction rate. The resulting hydrogen is assumed 1.38 to be uniformly distributed in the drywell. Figures 6.2-72D 1.39 and 6.2-72E show hydrogen generation rates and integrated values as a function of time following the accident. 1.40 Short-Term Ox en Source 1.42 The only source of air addition to primary containment, is 1.43 the operation of relief valves inside the primary containment. These relief valves are part of the breathing 1. 45 and service air systems, and. are normally isolated during 1. 47 reactor operation. Due to high temperature following a AVOCA 1.48 inside primary containment, a portion of these systems (inside primary containment) becomes pressurized and 1.50 relieves pressure by expelling about 126 standard cu ft of air into the primary containment.
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 1. 51 storage of portable air packs for breathing. The operating 1.52 procedures would have appropriate controls for the use of portable ai r packs.
Amendment 6.2-79a ch1217718f-14dc 09/06/84 155
Nine Mile Point Unit 2 FSAR THIS PAGE INTENTIONALLY BLANK Amendment 6. 2-.79b.
chl217718f-14dc 09/06/84 155
Nine Mile Point Unit 2 FSAR The automatic depressurization system (ADS) valves are 1 12 nitrogen operated; therefore, operation of these valves will
~
not result in addition of oxygen in the primary containment. 1. 13 The short-term oxygen source has not been considered in the 1. 15 oxygen concentration evaluation, as it is very small. 1. 16
~Eon -Term H dro en/Ox~en Generation 1. 18 Hydrogen and oxygen are produced by decomposition of water 1. 19 due to absorption of the fission product decay energy 1.21 immediately after a AVOCA. Generation of hydrogen and oxygen 1.22 due to radiolysis of core cooling water is an important 1.23 factor in determining the long-term gas mixture composition within the primary containment. A fission product 1. 24 distribution model as outlined in Reg'ulatory Guide 1.7 is used to calculate hydrogen/oxygen generati'on rates. The 1. 26 incore radiolysis (due to core gammas) contributes hydrogen and oxygen to the drywell, and radiolysis due to fission 1.27 products contributes hydrogen and 'xygen directly to the 1.30 suppression chamber and the drywell atmospheres. The 1.31 division of hydrogen and oxygen between the suppression chamber and the drywell depends upon the fraction of water 1.32 holdup on the drywell floor and water in the reactor'essel.
Hydrogen can also be formed by corrosion of metals and 1.39 decomposition of organic materials in the primary containment. The significant portion of this source is from 1. 40 the corrosion of zinc, which is included in the l. 41 temperature dependent hydrogen production rate is based analysis'he
exposed to sprays is shown in Table 6.2-59D. The surface 1. 45 area used in the analysis is about 15 percent higher than the tabulated values. The corrosion of aluminum in 1.46 demineralized water is very small. The Griess and 1. 47 data suggest the hydrogen production rate to be between 1. 48 Creek'est 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 1. 49 rate is 4x10 'td.
mainly due to 285'F and 212 F water temperature, the average cu ft of H~ per sq ft per hour.
Considering the aluminum surface area'irectly exposed to 1.50
- 1. 51.
the spray environment and the above Hz generation rate, a 1.52 total of 125 SCF of hydrogen would be evolved within 20 days 1.53 following ,a I.OCA. This being very small compared to other 1.54 sources of hydrogen, Al corrosion and associated hydrogen 1.55 production is ignored in the analysis.
Figures 6.2-72D through 6.2-72G .show hydrogen and oxygen 1. 56 generation rates and integrated values. The . quantity of 1. 57 Amendment 6.2-80 chl217718f-14dd 09/05/84 155
Nine Mile Point Unit 2 FSAR initially reactor coolant hydrogen system is contained negligible; hence, it within the is neglected.
6.2.5.3.2 Accident Description 2 1
~
Following the postulated recirculation suction line double- 2.2 ended rupture, the metal-water reaction begins in the core region and produces hydrogen immediately. The reaction is 2.7 assumed to last 2, min, during which 0.7 percent of the active zircaloy .
fuel: cladding reacts.. The radiolysis of 2;8 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.
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 combustible gases in the drywell and the suppression The 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 1.13 will permit operation of the recombiner. The or recombiner 1.15 system is manually activated when oxygen hydrogen concentration reaches 4.5 percent. The recombiner system 1.16 takes suction from the primary containment atmosphere, recombines the hydrogen and oxygen to form water vapor, and 1.18 returns the exhaust 'to the suppression chamber. This 1.19 results j.n a small pressure buildup in the suppression chamber that causes the opening .of the vacuum breaker valves 1.21 between the drywell and suppression chamber. As a result, 1.22 the flow of the gas mixture from the suppression chamber to the drywell is established. This arrangement. of recombiner 1.23 suction and discharge promotes mixing of the two volumes in the primary containment.
6 '.5.3.3 Analysis 1.25 .
Based on the preceding hydrogen and oxygen generation 1.26 sources and the accident description, the oxygen and 1.28 hydrogen concentration in the drywell and suppression chamber is obtained as a function of time. To calculate the 1.29 redistribution of the hydrogen and oxygen between the drywell and suppression chamber, a two-region computer model 1.30 of the primary containment system is used. This model takes 1.32 into consideration hydrogen and oxygen generation from the metal-water reaction and ~radiolysis. The calculation 1.33 determines the inventory, partial pressure, and mole fraction, of each atmospheric constituent in both regions as 1.35 a function of time.
Tables 6.2-58, 59, 59C, and 6.2-59D present the parameters 1.36 used in the analysis of the oxygen and hydrogen buildup 1.37 within the primary containment. Figures 6.2-72H and 6.2-721 1.39 present hydrogen and oxygen concentration transients in the primary containment, assuming only one recombiner is 1.40 operating. The recombiner is required to be functional 1.41 approximately 2.75 days after the design basis accident.
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 1.17 This system will be tested reactor power operation.
periodically 'to assure that it whenever required.
'ill" operate correctly Preoperational tests of the CGCS are 1.19 1.20
'.22 conducted during the final stages of plant construction prior to initial startup. These tests assure correct 1.23 functioning of all controls, instrumentation, recombiners, piping, and valves. System reference characteristics such 1.24 as pressure differentials and flow rates are'ocumented
~
during the preoperational tests and will be used as base 1.25 points for measurement. in subsequent operational tests'uring normal operation, the recombiner system piping, 1.26 valves, instrumentation, wiring, and other components can be 1.27 inspected visually at any time, since they are outside the primary containment. Further information may be found in 1. 28 Chapter 14.
- 6. 2. 5. 5 Instrumentation Requirements 1.30 1.31 Safety-related instruments, and controls are provided for 1.32 automatic and manual control of the hydrogen recombiners. 1.33 The controls and monitors described below are located in the 1.35 main control room. The control logic is shown on 1 '6 Figure 6.2-72K.
Instrumentation requirements for the primary containment l. 37 purge. system and the SGTS portions ef the CGCS are described 1.39 in Sections 9.4.2.5 and 6.5.1.5, respectively.
1.41 The hydrogen recombiner inlet and outlet isolation valves 1.42 close automatically on. a LOCA or manual isolation signal and 1.44 can be opened manually during a LOCA by means of the associated hydrogen recombiner LOCA override keylock switch. 1. 45 Amendment 6.2.-82 ch1217718f-14di 09/05/84 155
Nine Mile Point Unit 2 FSAR 6.2.7 References 1. 13
- 1. Models used in LOCTVS A Computer Code to Determine 1.15 Pressure and Temperature Response of Vapor Suppression Containments Following a Loss-of-Coolant Accident, 1. 16 Topical Report SWECO 8101, 1981. 1.17
- 2. Maximum Flow Rate of a Single Component Two-Phase 1.18 Mixture, APED-4378, October 25, 1963.
- 3. Sharma, D. F. Technical Description Annulus 1.19 Pressurization Load Adequacy Evaluation, NEDO-24548, January 1979. 1.20
- 4. NUREG/CR-2812 (January 1984), The Relative Importance of 1.22 Temperature, pH 'and Boric Acid Concentration. on Rates of Hz Production from Galvanized Steel Corrosion. 1.23
Amendment .6. 2-89 ch1217718f -14d j 09/05/84 155
Nine Mile Point Unit 2 FSAR TABI E 6. 2-59 PLANT PARAMETERS USED IN POST-DBA COMBUSTIBLE GAS CONCENTRATION ANALYSIS Reactor power 3, 467 MW 1.16 Drywell free volume 303,418 ft~ 1.25 r
Suppression chamber free volume 192,028 ft~ 1.27 (at high pool water level) 1.28 Initial drywell pressure 15.45 psia 1. 30 Initial drywell temperature 135 F 1.32 Initial drywell relative humidity 40% 1.34 Initial suppression chamber 15.45 psia 1.36 1.37 pressure Initial suppression 90 F 1.39 chambe r tempe r a tur e 1.40 Initial suppression chamber 100/ 1. 42 relative humidity 1. 43 Weight of- zircaloy in core 93,246 1bm 1. 45 (active fuel) 1. 46 Zircaloy reaction with steam 0.7% 1.48 Duration of reaction 120 sec 1.50 Fraction of water. in drywell 5.9% 1.52 J.
and reactor vessel 1.53 Downcomer submergence at high pool 11 feet 1.55 water level 1.56 Vacuum breaker set, point 0.25 psid 1.58 Initial Oz concentration 4 volume percent 2. 2 Recombiner- capacity 150 scfm 2.4 p
Amendment 1 of 2 ch1217718f-14df 09/05/84 112
Nine Nile Point Unit 2 FSAR TABLE 6.2-59 (Cont)
Recombination efficiency ~100% 2.6 Temperature transient for Figure 6.2-8 2.9 primary containment 2.10 (recirculation suction 2.11 line DER) 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 4.0 x.10 "
j Corrosion Rate
~SCF~ft~-hr (constant)
Applicable Temperature Up to 285~F 1.14 1:15 1 ~ 16
- 1. 18
-5113.25 1.20 Zinc 0.6764 exp (460 + T) 119.12 F 5 T 5224.06 F 1.21
-23416.67 1.23 2.8245 x 10 exp 460 + T) 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 ,Surface Area gft~}
Aluminun 650 41, 500 Galvanized steel 6, 968 58,540',400 Zinc primer 230 Amendment 1 of.l chl217718f-14dh 09/05/84 112
4 CJ tO SuP . Chas)kI Z RadI OtP SI 5 Dyy~e l t R~d'>> n1y~I~
C)
LLI CD lY CI
'2i Yl c.
x:
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 Wo+~ l C3 I
CK J) ~~Q, I I 4J P achi oipiS 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
U Ll (0
K Cl I
CK CY bl g
hl Q>y~ Q l I C)
PAd i ol Sic C3 4J I
CT CY hl I
0 04 TIt1E AFTER ACCIOENT (SECDNOS)
FIGURE 6.2-72G TNTDhhTED OXYOKN ODINATION FOLLDMINO'eh
m/o Qf(.os Lal (J
X C3 tJ
~4 c>
5~~ af Supp-
~ceo 2
~ea 3
~sea
+
~eoe G
~~C w~ee-7
~
TINE RFTER RCCIOENT (SE&B%%)
bAQS FIGURE 6.2-72H OXTQEN COCOIThhTlOH FOLLOMlHO OOh
,>r lt I
UJ LJ UJ 0
kQ(~M Irg n X Sv pP C3 I C 4 A'h k; . ~
CC Q
I 4J C3 S LIP P.
'>y~c l UJ C4.'i rh Q ft$
C3 kg<,ornbin g S t-ac I-s lY C3 Bn ye 8 oq +y v/0 X: )ty,l >n. r. of Z.l 5 dog t92866 ~960 3
~866- ~~G
~f499 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 Location of valve length of Inside/ Pipe Con- Valve~ is)
Pene- FSAR Outside taanment to Potential Isola-GDC or Arrange- Primary Outermost Type Bypass Position tion Closure Power tration System Reg. ESF Size ment Contain- Isolation Test Leakage Oper- Actuator Mode Normal Post- ~power Q) Signal Time Source li . ~D Guide ~S stem Fluid ~in) ~Fi ure( ~ > ment Valve ((> Path(>> Number Tel> e ator ~Pr i mar secondary Shutdown Accident Failure c+> (5 e> ( T> Note.
Z-1A Main steam Line A 55 No Steam 26 6.2-70 Inside 5 '2w C Yes 2MSS+HYV6A Ball HYV Hydraulic N/A Open Closed closed closed B,C,D, 3 to N/A 8 Sh. 1 outside C 2MSS+HYV7A Ball HYV to open; E,P,T, 5 sec spring Rr RM Main steam to close Line A Outside 2MSS+AQV93A Globe AOV Closed Closed closed Closed N/A RM N/A drain line Outside 2MSS4MOV208 Globe MOV Closed Closed Closed Closed FAI BrCro ~ N/A FrErPr Z-1P Main steam 6.2-70 5'-2" Line 8 55 No Steam 26 Inside C Yes 2MSS+HYV6B HYV Hydraulic N/A Open closed Closed closed B,C,D, 3 to N/A Sh. 1 Outside C 2MSS+HYV7B Ball HYV to open; 5 sec N/A spring T,RM Main steam to close Line Outside 2MSS*AOV938 Globe ACV Closed Closed Closed Closed RM N/A B
drain line Z-1C Main steam 5'-2" Line 55 No Steam 26 6. 2-70 Inside C Yes 2MSS*HYV6C Ball HYV Hydraulic N/A Open Closed Closed Closed B,C,D, 3 to N/A 8 C Sh 1 outside C 2MSS*HYV7C Ball HYV to open; 5 sec Main steam spring R,RM Line C Outside 2MSS( AOV93C Globe AOV to close Closed Closed Closed Closed RM N/A drain line Z 1D Main steam Line 55 No Steam 26 6.2 70 Inside 5'-2" C Yes 2MSS+ HYV6D Ball HYV Hydraulic N/A Open Closed Closed closed B C D 3 to N/A 8 D
Sh. 1 Outside C 2MSS+HYV7D Ball HYV to open; E P T 5 sec Main steam spring R,RM Line D Outside 2MSS+AOV93D Globe Aov to close Closed Closed closed Closed RM drain line Z-2 Main steam dram lane 55 Steam 6 6. 2-70 Inside 1'-0" C Yes 2MSS*MOV111 MOV Elec. Manual Closed Closed Closed FAI B,C ~ D Div Div II I
Sh. 2 outside C 2MSS+MOV112 MOV Elec Manual Closed Closed Closed FAI Z 3 Spare TI 25' APERTURE@ of 3>r CARD
Docke", ~tt Contre Nine Mile Point Unit 2 FSl
'E<~<
. -'eument
-"'U'-KETFILp TABLE 6. 2-56 (Cont)
Location of valve Length of Inside/ Pipe Con- Valve(e)
FSAR Outside tainment to Potential Pene- or Arrange- Primary Isola-tration II .
System GDC Reg.
Guide ESF
~Sstem Size Fluid ~in ment
~iciuure( ( )
Contain-ment Outermost Isolation Valve Type Test
())
Bypass Leakage Path(z)
Oper-ator Actuator Mode Normal Position Post- Power))
'ion Signal Closure Tzme Power Source Number Tyi)e ~P (3) Shutdown Accident Farceur (5 t)) (7) Note
~P-4A Feedwater 55 No Water 24 6. 2-7P Outside 2P 1<< Swing /7 C 2FWS<<AOV23A AOV Pneumatic line A to RPV Sh 3 Inside 0 ~
0 It W~) 2FMS*V12A Check Swing Flow Open Closed Closed Closed Reverse The time,N/A flow it takes N/A Open Closed Closed Closed for one check valve volume to pass through the valve Z-4B Feedwater 55 No Water 24 6. 2-70 Inside OP-ott Swing N/A line to N 2FWS'4V]2B Flow N/A Open Closed Closed N/A Reverse The time A RFV Sh. 3 m)
Outside 2 '1<< 2FMS*AOV23B Check Swing AOV Pneumatic N/A Open Closed Closed N/A flow it takes for one Check valve volume to pass through the valve Z-5A RHS Pump A 56 Yes Mater 24 6. 2-70 Outside 5'-6<< Tricen- Elec suction from No 2RHS+MOV1A MOV Manual Open Closed Open FAI Div I suppression Sh- 4 tric RM 45 butter- 13 pool fly Z-5B RHS Pump B 56 Yes Water 24 6.2-70 Outside 20P-g<< Tricen-suction from 2 RHS*MOV1 B MCV Elec. Manual Open Closed Open FAI Div II suppression Sh. 4 tric RM 45 Czg) butter- 13 pool fly Z-5C RHS Pump C 56 Yes Mater 6.2-70 Outside gt-gpt Tricen-suction from 24 Sh.
C No 2RHS<<MOV1C tric MOV Elec. Manual Open Closed Open FAI RM 45 Div II suppression Lz'f ) butter- 13 pool fly Z-6A RHS test line 56 Yes Mater 18 6 2-70 Outside 9 ' 5/16<<C No Tricen- Elec. Manual Closed Closed Loop B to sup- Sh 6 1 2RHS+MOV30B tric MOV Open FAI RM
- Div I pression pool (zy ) butter- 15 fly APERTURE 2 of~
CARD
$ So9mumeu-oa
c X4c2( I Nine Mile Point Unit 2 FS)
TABLE 6 2 56 (Cont)
Location IL@
of valve Length of Inside/ Pipe Con- Valve
('I1 FSAR Outside tainment to Potential Isola-Pene- GDC or Arrange- Primary Outermost tration System Size Contain-Type Bypass Position tion Closure Power Reg. ESF ment Isolation Test Leakage Oper- Actuator Mode Normal Post- Pouter l@ Sianal T).me Source No.
~dHS Guide ~s stem Fluid ~in ~Fi ure(( > ment Valve (() Path(z) Number T~oe a tor P .
d ~dd ( )> shutdown Accident Failure (n) ( z) Noti Z-68 test line 56 Yes Water 6. >-70 Outside Tricen-Loop A to sup-18 Sh. 6 9 ~
3 dd 2RHS+MOV30A tric MOV Elec. Manual Open Closed Closed FAI Div II pression pool butter- 15 fly Z-7A RHS containment 56 Yes Water 6. 2-70 Outside 1st 3n spray Loop A Sh. 7 2RHS+MOV33A Globe MOV Elec. Manual Closed Closed Open FAI G 23 Div I to suppression pool Z-7B RHS containment Yes Water 70 Outside spray Loop B 56 4 6 Sh.
2 7
4 ~
6 II C No 2RHSnMOV33B Globe MOV Elec. Manual Closed Closed Open FAI G 23 Div II to suppression pool 7/( 15 Z-BA RHS containment 56 Yes Water 16 '-70 Outside 2 ~ Q II spray Loop A 6
Sh. Outside C No 2RHs<<Mov25A Gate MOV Elec. Manual Closed Closed Open FAI RM 89 Djv I to drywell 8
c<<)
Z BB containment Water 6. 'r-70 RHS spray Loop B 56 Yes 16 Sh. 8 Outside Outside 2 ~ 3 n Nog ~) 2 RHS< MOV25 B Gate MOV Elec. Manual Closed Closed Open FAI RM 89 Div II to drywell 15 Z 9A RHS/LPCI 55 Yes Water 6-:-70 Outside 7)
LooP A to RPV 12 Qn C No~ g 2RHS+MQV24A Gate MQV Elec Manual Closed Closed Open FAI RM 19 5 Div I Sh. 9 Inside C 2RHS+AOV16A Check ACV Process N/A Closed Closed Open Closed Reverse flow
~-o 15 Z 9B RHS/LPCI Loop B to RPV 55 Yes Water 12 6. 7-70 Sh. 9 Outside Inside 6'-6<< No Wi 2RHS*MOV24B Check AQV Elec Process Manual N/A Closed Closed Closed Closed Open Open FAI Closed RM 19.
Reverse 9 g-o 5 Div II C 2 2RHS+AOV16B N/A 1 flow 15,11 Z-9C RHS/LPCI Loop C to RPV 55 Yes Water 12 6. 7-70 Sh.
Outside Inside 6)-6>> C No 2RHS<<MQV24C Check Elec.
Process Manual Closed Closed closed Open FAI RM 19 5 Div II 9 C g'g) 2RHS<<AOV16C AQV N/A Closed Open Closed Reverse S 9'O N/A )
flow 15,1) )
TI 3 off' APERTURE CARD 509MQ962 -cS
()nt>))1
+o P//o g <
REGULA,, )- -on Do Nine Mile Point Unit 2 FS>
" '-7'Fil,<
TABLE 6. 2 56 tcont)
Location of valve Length of (g>
Inside/ Pipe Con- Valve FSAR Outside tainment to Potential Isola-Pene- GDC or Arrange- Primary Outermost T'I/pe Bypass Position tion closure Power tration System Reg. ESF Size ment Contain- Isolation Test Leakage Oper- Actuator Mode Normal Post- PowerCR signal Time Source No
~OHS Guide System Fluid ~in ~Fi ure<<> ment Valve ((> Path(e> Number TPfRe ator ~P secon~dar Shutdown Accident Failure (t ) ( >) Not(
Z-10A shutdown 55 No water 12 6.2-70 Outside 6 ~ pn C 2RHSnNOV40A Globe MOV Elec. Manual Closed Open Closed FAI A,L,MD return Loop A Sh. 13 RM 25 Div I to reactor re- Inside 2RHSnAOV39A Check ACV Process Closed Open closed Closed Reverse 0 Div I circ Loop A flow @S RHS shutdown 55 No Water 2 6. 2-70 Inside 2RHSnNOV67A Globe MOV Elec. Manual Closed Closed Closed FAI A ~ L,M ~ Div I cooling return Sh. 13 RM line inboard valve bypass line Z-10B RHS shutdown 55 No Water 12 6. 2-70 Outside 6I pn C No , 2RHSnNOV40B Globe MOV Elec. Manual Closed Open closed FAI A,L,NP 25 Div I return Loop B Sh. 1] (z't > RM to reactor re- Inside 2RHSnAOV39B Check AOV Process N/A Closed Open closed Closed Reverse&5-(3 Div I circ Loop B flow RHS shutdown 55 No Water 2 6. 2-70 Inside 2RHSnMOV67B Globe MOV Elec. Manual closed closed Closed FAI A,L,M, 9 Div I cooling return Sh. 13 RN line inboard valve bypass line Z-11 RHS shutdown 55 No Water 20 6.2-70 Outside 6 ~ pn 2RHsnMOV1 13 . Gate MOV Elec Manual Closed Open closed FAI A,LE iM, 27 Div I supply from Sh. (Zg ) RM reactor recirc 14 Inside 2RHsnMOV112 Gate MOV Elec. Manual Closed Open Closed- FAI A,L,M, 27 Div II RN Inside C HSnRV152 Relief N/A Auto N/A Closed Closed Closed Closed N/A N/A N/A
/ e.s FAI Div III Z 12 suction water Gate Elec. Manual Closed Closed Open RN 16 CSH from suppres-sion pool 56 Yes 20 6. 2-70 Sh. 5 Outside 2 ~
2 II C C ~go) 2CSHnMOV118 MOV Z-13 CSH test return 56 Yes Water 12 6. 2-70 Outside 5Q ~ Qn No 2CSHnMOV111 Globe MOV Elec. Manual Closed Closed Closed FAI 8, F,RM STD Div III to suppression Sh. 15 (z'f ')
HPCS min flow Yes Water 4 Outside 45I 6n 2CSH*MOV105 Gate MCV Elec Manual Closed Closed Closed FAI RN STD Div III bypass TI 4 of 2(0 APRRTURI,'ARD
PLV>Sel n n>J +l -'//r> Nine Mile Point Unit 2 Fs!
// &Ay f2~~
WRy D'OC ocun>enf TABLE 6. 2-56 (Cont)
ET FQ@
location of valve Length of (4)
Inside/ Pipe Con- Valve Outside tainment to Potential Isola-Pene-tration No
~DDC System Reg.
Guide or
~s ESF stem Size Fluid ~in FSAR Arrange-ment Fixture<<>
Primary Contain-ment Outermost Isolation Valve Type Test (I >
Bypass Leakage Path<z> Number Oper-ator Actuator Mode secon~dar Normal
( >>
Position Shutdown Post-Accident P~ove Failure lO) tion Closure Signal C~ >
Time CS II>
Power Source t>> Mott Z-14 CSH to RPV 55 Yes water 12 6. 2-70 Inside 2CSH~AOV108 Check AOV Process Air Closed Closed Open Closed ,Reverse N/A flow 1'3 Sh. 9 Outside 2CSH~MOV107 Gate MOV Elec. Manual Closed Closed Open FAI RM 27 Div III Z-15 CSL suction 56 Yes Water 20 6. 2-70 Outside 1 I 8II 2 CSL+ MOV1 1 2 But ter- MOV Elec. Manual Open Open Open FAI RM 90 Div I from suppres-sion pool Sh. 4 fly 13,-
Z-16 CSL to RPV 55 Yes water 12 6. 2-70 Inside C 2CSL~AOV101 Check AOV Prccess Air Closed Closed Open Closed Reverse N/A N/A 11 Sh. 10 flow 13 Outside 2CSL+MOV104 Gate MOV Elec. Manual Closed Closed Open FAI RM 637 Div I fCa Z-17 ICS suction 56 Yes Water 6 6. 2->0 Outside 9 II 2ICSI'MOV136 Gate MOV Elec. Manual Closed Closed Open FAI RM 30 120VDC from suppres- Sh 5 sion pool Z-18 ICS minimum 56 Yes Water 2 6.2- 0 Outside 6 II 2ICS+MOV143 Globe MOV Elec Manual Closed Closed Open FAI RM 5 120VDC Z-19 flow to sup-pression pool ICS turbine exhaust to suppression 56 Yes Steam 12 Sh.
Sh. 12
'1
- 6. 2 70 Outside 2.75'o C No gz'l)
(z~)
2ICS~MOV122 Gate MOV Elec. Manual Open Open Open FAI RM 61 120VDC pool ICS turbine 56 Yes Steam 1 1/2 6. 2-70 Outside 2ICS4MOV164 Globe MOV Elec. Manual Open Open Open FAI F~RM ~ H 10 exhaust S11. 12 f vacuum breaker Z-20 Spare No 3/4 TI APERTURE CARD 5 of+
0 '//o Nine Mile Point Unit 2 I Control P /<'md~of 13ocun)en 13ate- ~/ Y llpCKET TABLE 6. 2-56 (Cont)
Location of valve Length of Inside/ Pipe Con- EEOC"'osition Pene- FSAR Outside tainment to Potential GDC or Arrange- Primary Isola-tration System Reg, ESF Size Contain-Outermost Type Bypass No. Guide ~sstem Fluid ment Isolation Test Leakage Oper- Actuator tion Closure Power
~in ment Mode Normal Post- Signal Z-21A
~0team to ICS Yes Steam
~F'.2-70
- Valve ( <) Path(n) Number ator pPrimary ~dd (s) Shutdown Accident Powers Failure t n)
<4)
Time
<s e)
Source =.
Not turbine and RHS 10 Outside 9N C 2ICS*MOV121 Gate Inside, NOV Elec. Manual Closed heat exchangers Sh. 16 1 II C 2ICSsNOV128 Gate MOV Elec. Manual Open Open closed Open Open
~ FAI FAI KERN 14. 5 Div I K,RN 14 5 ICS turline Steam Inside steam supply 2ICS*MOV170 Globe .'OV Elec. Manual Open Closed Open FAI K,RM 10.5 bypass to inboard isolation va lve Z-218 Spare Z 22 ICS to RPV 55 Yes Water 6 6. 2-70 Outside 4.25" No 2ICsd'AOV156 Check AOV Process Air Closed Closed Sh. 17 (29) Open Closed Reverse 5 120VDC Inside flow 2 ICS*AOV157 Check AOV Process Air Closed Closed Open Closed Reverse 5 flow Z-23 WCS from supply RCS 8 RPV 55 No Water Water 8
8
ELec.
Manual Manual Open Open Open Open Closed Closed FAI FAI U, B,RM U,Bd Jd 14 14 Div Div I
I Z-24 Spare Z-25 RDS lines to Yes RPV See Note 17 53 Insert Water 1 N/A Outside
.53 Withdrawal 3/4 125 125'utside,
~
Z-26 RDS lines to Yes RPV 39 Insert See Note 17 TI 39 Withdrawal Water 1 3/4 N/A Outside 125'utside
(~>) APgg~
Z-27 RDS RPV lines to 54 Insert 54 Withdrawal Yes Water 1 3/4 N/A 125'/0 125'utside 125'utside u/0 C2~)
See No e 17 CARD 6 of/
f 617i5<PI Nine Mile Point Vnit 2 FSA KD Y/C3 Yo F2~ TABLE 6 2-56 (Cont) 1e'7 337~P location T't77@
of valve Length of Valve(g)
'DC 125'25'ee Inside/ pipe Con-FSAR Outside tarnment to Potent1.al Isola-Pene-tration No.
Z-28
~tt RDS System lines to Reg.
Guide or
~S ESF size stem Fluid ~in)
Yes Arrange-ment Primary Contain-ment Outermost Isolation Valve Type Test
<13 Bypass Leakage Path<<'> Number Oper-ator
'ctuator Primary Mode secon~dar Normal (31 Position Shutdown Note 17 post-Accident PowerWIO)
Failure tion Closure-Signal Time Power Source Note RPV 39 Insert Water 1 N/A Outside 39 Withdrawal 3/4 Outside Z-29 SLCS to RPV 55 Yes Boron solu-1 1/2 Inside Outside No 2SLS'tv10 Check N/A Process N/A Closed Closed 'losed flu~
2 S Lse'OV5 A Stop MOV Elec. Manual Closed Closed Ru/CN' tion check globe Outside 2SLS+MOVSB Stop MOV Elec. Manual Closed Closed 0/A t/1/A check y/aw globe C.BOA Z-30A Spare Z-30B Spare i/7' TIP drive guide '-3/4" Z-31A tube to RPV 57 No Note 19 1 1/2 6. 2-70 Sh. 1'j Outside outside 1
"(3) N/A N/A Ball shear SCV N/A Elec.
N/A Elec.
N/A Closed Open Closed Open Closed Open Closed Open iten weP N/A 120
'45'p 18, 19
'e773C Z-31P TIP drive guide 57 No Note 1 1/2 6. 2-70 Outside 1 ~ -3/4" N/A Ball SOV Elec. Elec. Closed Closed Closed Closed N/A 1 20.
tube to RPV 19 Sh. 19 Outside (3/) N/A Shear N/A N/A N/A Open Open Open Open 1s-&
QC 18,19
~73C.
Z-31C TIP drive guide 57 No Note 1 1/2 6. 2-70 Outside 1 ~ -3/4w No N/A Ball SOV Elec. Elec. Closed closed Closed Closed N/A 1 20-tube to RPV 19 Sh. 19 Outside ( ) N/A Shear N/A N/A N/A Open Open Open Open z-s QC 18 19
~ VP 1
~
V+ C, Z-31D TIP drive guide 57 No Note 1 1/2 6. 2-70 outside e-3/4'e No N/A Ball SCV Elec. Elec Closed Closed Closed Closed N/A 120-( 37) 1 tube to RPV 19 S11 19 Outside Shear N/A N/A N/A Open Open . Open Open lz-> lse19 1773 ct Z-31E 'IIP drive guide 57 No Note 1/2 6.2-70 Outside 3/4u No Ball SOV Elec. Elec Closed Closed Closed Closed N/A 20-.
( 31) 1 ~ 1
, 1 tube to RPV 19 Sh. 19 Outside Shear N/A N/A N/A Open Open Open Open 1~ 18 ~ 19 Z-32 Nz purge to ' oeCQ
~os Closed Closed Closed
+w VfhC No Nz 1 1/2 Outside 1
~ 0 ee No (3 I) N/A Check Simple N/A N/A N/A N/A.
TIP index 57 check ea ches t77 K u.3'-o" cLC3 c.D l2C3 VAC mechanism f 31) C sc,v U-13'lobe Z-33A CCP supply to 56 No Water 4 6. 2-70 Inside C No ( 3I) 2CCPe'MOV94A MOV Elec. Manual Manual Open Ooen Closed Closed FAI B~ F,RM 2.z Div Div I II 76 RCS Pump A Sh 20 Outside 7 ~
0 ee C 2CCPe'MOV Globe MOV Elec. Open Open FAI B,F,RM 23
~7 7R 7 0
l)o
'<y Q IZ L//5< (
c.,
unf/b/ P- Y/
of /) one Mzle Point Unxt 2 FSA1 Oppugn/en6 TABIE 6. 2-66 (Conti Location of valve I.ength of Inside/ Pipe Con- ,ta)
FSAR Outside tainment to Potential Isola-Pene- GDC or Arrange- Primary Outermost Type Bypass Position tion Closure Power tration System Reg. ESF size ment Contain- Isolation Test Leakage Oper- Actuator Mode Normal Post- Powe Io)Signal Time Source No. D Guide ~sstem Fluid ~in ~iciure<L4 ment Valve CL4 Path>>>> Number ator secon~dar Shutdown Accident Failure (s >>) C z1 Notes Z-33B CCP to Rcs 56 No Water 4 6. 2-70 Inside C ..No (3/) 2CCP>>MOV94B Globe MOV Elec. Manual Open Open closed PAI BE F,RM Div II Pump B Sh. 20 Cutside 7 ~
0 II C 2CCP>>MO 4B Globe MOV Elec. Manual Open Open closed FAI BrF ARM 23 Div I i(A Z-34A CCP return from 56 No Water 6. 2-70 Inside C ~No (3/) 2CCP>>MOVE) Globe MOV Elec. Manual Open Open Closed FAI.. B~ F,RM 23 Div II
- Z-348 RCS Pump A CCP return from, 56 No Water 4 Sh. 21
- 6. 2-70 Outside Inside 7 ~ Off r
C C No (3/)
2CCP>>MO~/I~-Globe
/6 IB 2CCP>>MO 1 Globe
. MOV MOV
-.E 1 ec=,.-
Elec.
. Manual Manual cI Open Open Open Open
-,;;;Closed r.-.
Closed FAI=,=
/'3'3 Dim(I Div II RCS Pump B r, ,
Sh. 21 Outside 7 ~ 0 II C 2CCP>>MO - Globe MOV Elec. Manual Open Open Closed FAI BE F,RM Div I
/58
--Z-35 Spare
.Z 36 Service air to 56 No Air 6.2-70 Outside 7 II NQ ( 3/) 2SAS>>HCV161 Globe Manual Manual N/A closed Closed Closed closed LMCrLC N/A Div I drywell ~
Sh. 22 Inside 2SAS>>HCV163 Globe Manual Manual N/A Closed Closed Closed Closed LMC ~ LC N/A Div II Z 37 Breathing air 56 No Air 6. 2-70 Outside C No (3 /) 2AAS>> HCV1 3 4 Globe Manual Manual N/A Closed Closed Closed Closed LMC ~ LC N/A Div I to drywell Sh. 22 Inside C 2AAS>> HCV1 3 6 Globe Manual Manual N/A Closed Closed Closed Closed LMC,LC N/A Div II Z-38A RDS to recirc pump A seal 55 No Wa ter 3/4 6. 2-70 Sh. 23 Inside Outside <5I
'9'i 2RCS>>V60A Check N/A Plow N/A Open Closed Closed N/A Reverse flow N/A On 2RCS>>V90A Check N/A Flow N/A Open Closed Closed N/A Reverse flow Outside 2RCS>>V59A Check N/A Flow N/A Open Closed Closed Reverse flow Z-38B RDS to recirc Pump A seal 55 No Water 3/4 6. 2-70 Sh. 23 Inside (5 ~
0 It No 'z9 2RCS>>V60B Check N/A Flow N/A Open Closed Closed Reverse flow
////i, N/A Outside 2RCS*V90B Check Flow N/A Open Closed Closed N/A Reverse r
flow outside 2RCS*V59B Check N/A Flow N/A Open Closed Closed N/A Reverse flow Z-39 Drywell floor drain tank 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 Div II I
vent line yes Z-40 Equi pme nt drains from 56 No Water 4 6. 2-70 Inside outside 4 ~
2 1/2>> C
(~o) 2DER*MOV119 2DER>>MOV120 Gate Gate MOV Elec.
Elec.
Manual Manual Open Open Closed Closed closed Closed FAI FAI B,F,RM B,E,RM 21.
21 3 3 Div Div II I
Sh. 24 C MOV drywell 8 of 24
Nzne Mrle Pornt Unzt
~eo 2 PSALM TABLE 6 2-56 (Cont)
.e~ >eh Location of valve Length of Inside/ Pipe Con- Valv FSAR Outside tainment to Potential Isola-Pene- GDC or Arrange- Primary Outermost TYPe Bypass Position tion Closure Power tration System Reg ESF Size ment Contain- Isolation Test Leakage Oper- Actuator Mode Normal Post- Powe Io3 Signal Time Source N .
Guide ~s stem Fluid ~in ~iciure< 1 I ment Valve C1> Path<>> Number T~e ator Secon~dar Shutdown Accident Fai lur < + 4 (5 60 Note!
II
~O-41 Reactor coolant 55 Water 3/4 6- 2-70 Inside C No 2 RCS4 SOV1 0 4 Globe SOV Elec. Closed Closed Closed Closed B,F,RM Div recirc to Sh. 25 Outside (2i Pv C 2RCS*SOV105 Globe SOV Elec. Closed Closed Closed Closed B,F,RM Div I sample cooler Z-42A Eire protection 56 No Water 2 6.2-70 Inside C 2FPW+SOV219 Gate . SOV Elec. N/A Closed Closed Closed Closed B,F,RM for reactor ch 26 Outside C (3l ) 2FPW*SOV21 8 Gate SOV Elec. N/A closed Closed Closed Closed B,F,RM recirc pump Z-428 Eire protection 56 'Water 2 6- 2-70 Inside C 2F PW+ SOV2 2 1 Gate -.SOV Elec. N/A closed closed Closed closed ELF,RN water for reac- Sh. 26 Outside C (3() 2FPWOSOV220 Gate SOV Elec. N/A Closed Closed Closed Closed B,F,RM tor recirc pump yes
( Z-43 Z-44A Drywell floor drains Capped spare 56 Water 6. 2-70 Sh. 27 In'side Outside C
C
~No (~c3 2DFR+MOV140 2 DFR*NOV139 Gate Gate MOV MOV Elec.
El ec.
Manual Manual Open Open Closed Closed Closed Closed
'FA FAI I B,F ~
B,F,RM RN 14 14-2 2
Z-448 Capped spare Z-44C capped spare Z-44D Capped spare Z-44E Service air to Air 6.2-70 Outside 5u No 2SAS+HCV160 Globe Manual Manual N/A Closed Closed/ Closed Closed LNC,LC N/A Div 'I 56 Inside C
(pi') 2SAS*HCV162 Open drywell Sh. 22 C Globe Manual Manual N/A Closed Closed/ closed Closed LMC,LC N/A. Div II Open Z-44F Breathing air 56 No Air 2 6. 2-70 Outside 5v No
() 2AAS~HCV135 Globe Manual Nanual N/A Closed Closed/ Closed Closed LMC,LC N/A Div I to drywell (3 Open Sh. 22 Inside 2AAS*HCV137 Globe Manual Manual Closed Closed/ Closed Closed LMC LC N/A Div II Open Z-45 Equipment drain Air 6. 2-70 Inside 2 DER I'OV 1 3 0 Globe MOV Elec. Manual Open Closed Closed FAI B F RN 9I ~ Div II 56 No 2 Outside C (3 O) 2DER4NOV131 Globe MOV Elec. Manual Open Closed Closed FAI B,F,RN 's.s Div I tank (2DER-'XK1) Sh. 27 C vent to drywell Z-46A CCP supply to 56 No Water 8 6-2 70 Inside 7' C 2CCP*NOV273 Gate MOV Elec. Manual Open Open Closed FAI B~ F, RN 40 Div Div II, I
Outside 2CCP~NOV265 Gate MOV Elec. Manual Open Open Closed FAI B,F,RN 40 drywell space Sh. 28 C cooler 9 of~
Tl APERTURE CARD 8~ho v~so ~Q~
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Dae I
Docnn)ant Y D'OCR'zz z) Nine Mile point Unit 2 F, TABLE 6 2-56 (Cont]
~F'ocation FSAR of valve Instde/
Outside Length of Pipe - Con-tainment to Potential Valve Z~3 I~so a-Pene- GDC or Arrange- Primary Outermost Bypass position tion closure Power tration Type System Reg. ESF Size ment Contain- Isolation Test Leakage Oper- Actuator Mode Normal Fost- Poweroa) Sional Time Source No
~d-468 Guide ~S stem Fluid ~in ment Valve ()) Path(z) Number TtfZ)e ator Primar ~Sd
- dt td Il 'd Eai lure (7) Not Capped spare Z-46C Eire protection water for con-See Note 20 Y) (Si) tainment hose reel standpipe Z-46D Capped spare Z-47 CCP return from drywell 57 No Water 8 6. 2-70 Sh. 28 Inside Outside 7 I 3II C C
No
~
2CCP~MOV122 2CCP<<MOV124 Gate Gate MOV MOV Elec Elec.
Manual Manual Open Open Open Open Closed Closed FAI FAI B~ FiRN BFFFRM 40 40 Div Div II I
space cooler I Z 48 Purge exhaust from drywell 56 No Ai r 14 6. 2-70 Inside (3(
2CPS+AOV108 Butter- AOV Pneu- Manual Closed Closed Closed closed BE F,RM 5 Div II Sh. 29 '1 fly matic outside C 2 C PS') AOV1 1 0 Butter- AOV Pneu- Manual Closed Closed closed Closed B,F,RM 5 Div I Ai(/g i fly matic Z-49 Purge to drywell inlet 56 No i'r 14 6.2-70 Inside 2CPS4AOV106 Butter- AOV Pneu- Manual closed Closed Closed Closed B F RM 5 Div II sh. 29 outside 2 CPS+ AOV1 0 4 fly Butter-matic Pneu- Manual Closed Closed Closed Closed B,F,RM 5 Div I AOV fly matic q)e Z-50 Purge to wetwell inlet 56 No Qr 12 6. 2-70 Inside 2 CPS'OV1 07 Butter- ACV Pneu- Manual Closed Closed closed Closed Div II sh. 29 C3 I) fly matic I
Outside 2CPS*AOV1 0 5 Butter- AOV Pneu- Manual Closed closed closed closed 8 F RM 5 Div fly matic Z-51 Purge exhaust from wetwell Arr 12 6 2 70 Inside "'(3>) 2" PS*AOV109 Butter- AOV Pneu- Manual Closed Closed Closed Closed B,F,RM 5 Div II Sh- 29 fly matic I
Outs).de 2CPSI'AOV111 Butter- AOV Pneu- Manual Closed Closed Closed 13 F RM 5 Div fly matic Z-52A Capped spare Z-52B Capped spare 1 P'e~
Z-53A Instrument air to ADS valve 56 No ir 1 1/2 6. 2-70 Outside 1 ~
0 If C M(3O> 2IAS~SOV164 Globe SOV Eleec. Open Open closed Closed Closed Closed B,F,RM B,F,RM CS N/A
+S O Div N/A I
Sh. 30 Inside 1 ~
0 II C 2IAS V448 Check N/A Process N/A Closed Open accumulators 2S N 10 ofM
'n
~KRONUR CARO
kef g g Contrui g d5vm y/
D dry
'LAI" RY o Document Nine Nile Point Unit 2 FSAR DOCKET FIDE TABLE 6- 2-56 (Cont)
Location of valve Length of Insider Pipe Con- Va~lve u7 FSAR Outside tainment to Potential Isola-Pene- GDC or Arrange- Primary Outermost Type Bypass Position +ion Closure Power tration System Reg. ESF size ment Contain- Isolation Test Leakage Oper- Actuator Mode Normal Post- PowerL<<3Siqnal Time Source No. ~d' Guide ~s stem Fluid ~in Ficiure ment Valve C1 4 Pathez4 Number ator ~dd Shutdown Accident Failure es az Notes Z-53B Instrument air 56 No 1 1/2 6.2-70 Outside 1
~ P If C ye 2IAS*SOV165 Globe N/A Elec. N/A Open Open Closed Closed BE F,RN Div I to ADS valve sh. 30 Inside 1
~ Pv C 2IAS*V449 Check Sov Process N/A Closed Open Closed Closed B,F,RM N/A N/A accumulators Z-53C Instrument air to NSRV accumu-56 No ~r 1 1/2 6. 2-70 sh. 30 Outside Inside 1
1
~
I P td 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 B,F ~ RM 4$ o Div Div I
II lator tank Z-54A Capped spare Z-55A Hydrogen recom-biner 1A supply to wetwell 56 Yes Air 6. 2-70 Sh. 31 Inside Outside C
C No 2HCS*MOV4A 2HCS~MOV1A Globe Globe MOV MOV Elec.
Elec Manual Manual Closed Closed Closed closed Open Open FAI FAI B,F ~ RN B,F,RM 18 5 18.5 Div Div I
I '2 Z-55B Hydrogen recom-biner 1B supply 56 Yes Air 6. 2-70 Sh 31 Inside Outside C
C No(3 ) 2HCS4MOV48 2HCSrNOV1B Globe Globe MDV MOV Elec.
Elec Manual Manual Closed Closed Closed Closed Open Open FAI FAI B,FrRM B,F,RM 18.
18.
5 5
Div Div II II 22 to wetwell Z-56A Hydrogen recom- 56 Yes Air 6. 2-70 Inside C No(+,i 2HCS*MOV6A Globe NOV Elec. Manual Closed Closed Open 'FAI Br'M 8 ',RM
C No 2HCS+NOV68 2HCS~MOV3B Globe Globe MOV MOV Elec.
Elec.
Manual Manual Closed Closed Closed Closed Open Open FAI FAI B,F,RN B,F,RN 18.
- 18. 5 5 Div Div II II 22 from drywell Z-57A Hyrdogen recom- 56 Yes Air 6.2-70 Inside C No ~ 2HCS*NOV5A Globe MOV Elec. Manual Closed Closed Open FAI BrFrRN 18 5 Div I biner 1A return Sh. 31 Outside C 2HCSl NOV2A Globe MOV Elec. Manual Closed Closed Open FAI B,F RM 18 5 Div I 22 from wetwell Z-578 Hyrodgen recom-biner return 56 Yes Air 6.2-70 Inside Outside C No
(
~ 2HCS+MOVSB 2HCS4MOV2B.
Globe Glube NOV Elec.
Elec.
Manual Manual Closed Closed Closed closed Open Open FAI FAI BE F,RM B,F,RN 18.5 18.5 Drv Div II II 1B Sh. 31 C MOV 22 from wetwell Z 58 Containment 56 No Air 2 6. 2-70 Inside C No (3 i) 2CPS+SOV122 Globe SCV Elec. Manual Closed Closed Closed Closed Closed B,F,RN 5 Div II I
purge to dry- Sh. 29 Outside C 2CPSrSOV120 Globe SOV Elec. Manual Closed Closed Closed B,F,RN 5 Div well TI 11 of APERTURE CARD
Rc viseg EGV A,pnv 7>ntun>enf
<.KP<7 Nine Mile Point Unit 2 FSAR FILp TABLE 6.2-56 (Cont)
I.ocation of valve I.ength of Insider Pipe Con-Pene- FSAR Outside tainment to Potential Valve( I>
GDC or Arrange- Primary Iso a-tration System Reg. ESF Size ment Contain-Outermost Type Bypass Position tion Closure No. Guide ~Sstem Fluid ~in Isolation Test Leakage Oper- Actuator Mode Normal Post-Power
~F> ure< > >
ment Valve I>> Path<a> Power Signal Time Source Z-59 ~0ontainment Number ator secon~dar Shutdown Accident Failure 3 (+ > 'ote, purge to wet-56 No Air 6.2-70 Inside we 11 Sh. 29 Outside C
C No (3/) 2 C Ps a SOV1 2 2CPS*SOV119 1 Globe Globe SOV SCV Elec.
Elec.
Manual Manual Closed Closed Closed Closed Closed closed Closed Closed B r F ~ RM B ~ F ~ RM Div Div I II Z-60A from dry-CMS well 56 No Air 3/4 6. 2-70 Inside C Sh. 32 Outside 2CMS*SOV6 lA 2CMS+SOV60A Globe SOV Elec. Elec. Open Closed Open Closed B, F, RM <1. 5 Div I Z 60B from dry-CMS well 56 Yes Air 3/4 6.2-70 Inside No Sh. 32 Cutside 2CMS+SOV24A 2CMS>SOV24C Globe SOV Elec Elec. Open Closed Open Closed B,ARM <1.5 Div I Z-60C CMS to dry- 56 No Air 3/4 6. 2-70 Inside well No L,) Globe Elec.
Sh. 32 Outside 2CMS+SOV63A 2CMS+SOV62A SOV Elec- Open Closed Open Closed B ~ F ~ RM <1.5 Div I Z-60D CMS to dry- 56 Yes Ai r 3/4 6. 2-70 Inside we 11 Sh. 32 Outside C " C~') 2CMS*SOV33A Globe SCV Elec. Elec. Open Closed Open Closed B F RM <1 5 Div I 2CMS+SOV32A Z-60E CMS from dry- 56 No Air 3/4 6. 2-70 Inside well Sh. 32 Outside No C3,q 2CMS 4 SOV6 2CMS+SOV60B 1 B Globe SCV Elec. Elec. Open Closed Open Closed B~ F,RM <1. 5 'iv II Z 60F from dry-CMS well 56 Yes Air 3/4 6. 2-70 Inside Sh. 32 Outside No C3 q 2CMSASOV24B 2CMS>SOV24D Globe SOV Elec. Elec. Open closed Open Closed B,F,RM <1.5 Div II Z-60G CMS to drywell No Air 3/4 6.2-70 Inside Sh. 32 Outside NoC5q 2CMS~SOV63B 2CMSWSOV62B Globe SOV Elec. Elec. Open Closed Open Closed Bi F RM <'1 5 Div II Z-60B CMS to drywell 56 Yes Air 3/4 6. 2-70 Inside Div II C No 2CMS*SOV33B Globe Elec. Elec.
Z-61A Sh. 32 Cutside C ) 2CMSASOV32B SOV Open Closed Open Closed B F Capped spare 3/4 Z-61B CMS from wet- 56 Yes Air we 11 3/4 6. 2-70 Inside Sh. 32 Outside NO/3 ~q 2CMS4SOV26A GloLe SOV Elec. Elec. Open Closed Open Closed BiFiRM <1.5 Div I 2CMS>SOV26C Z-61C CMS to wetwell 56 Yes Air 3/4 6.2-70 Inside No Globe Sh. 32 Outside ( 3 I) 2CMS>SOV34A 2CMS+ AOV35A SOV Elec. Elec. Open Closed Open Closed B,F,RM <1 5 Div I Z-61D Capped spare 3/4 Z-61E from wet-CMS 56 Yes Air 3/4 6.2-70 Inside well Sh. 32 Outside NQ (3 ~ 2CMS4 SOV26 B 2CMS*SOV26D Globe SCV Elec. Elec. Open Closed OPen C] osed B F RM <1 5 D>.v Il
'12 of ~
APERTURE CARD ggORQ5085Q-/0
Docket'o-Control g rf,'v~/~<+ <
ate Document Nine Mile Point Unit 2 FSAZ
OlRY DChCEET FILE TABLE 6.2-56 (Cont)
Location of valve length of Inside/ Pipe Con- Valvet+
Pene- or FSAR Outside tainment to Potential Isola-GDC Arrange- Primary Outermost Type Bypass Position tion Closure Power tration System Reg. ESF Sire ment Contain- Isolation Test Leakage Oper- Actuator Mode Normal Eost- P owe rr~ S zg na 1 Time Source No. Guide ~S stem Fluid ~in ~inure<<> ment valve (te Pathcer Number ator ~P ~Sd
--68 Capped spare 10 Z-69 Spare Z-70 Capped spare Z-71 Spare Z-72 capped spare 14 Z-73 RHS relief 56 No Water 6 6. 2-70 Outside 48'-6" No 2RHSeRV108 RV N/A N/A N/A N/A N/A N/A N/A None N/A N/A..
valve dis- Sh. 33 N
2RHSeRV20C charge to suppression pool Z-74 Capped spare Z-75 Capped spare Z-76 Capped spare Z-77 Capped spare 1 1/2 Z 78 Capped spare 1 1/2 Z 79 Capped spare 1 1/2 Z 80 Spent fuel 56 No Water 1 1/2 6. 2 70 Outside No 2SFCeV203 Globe Manual Manual Manual Closed Closed Closed Closed Closed N/A N/A pocl cooling Sh. 40 Inside q 2SFCA V204 Z-81 Capped spare 1 1/2 Z 82 capped spare zS TI 13 of Pf APERTURI CARD
3o-)i'toll Docket g f p, 3 Nine Mile Point Unit 2 FSAR P4 'ORY Df)CKQ, RT FILR TABLE 6. 2-56 (Cont)
Location of valve Length of Pene-tration No ~OtSystem
'DC Reg.
Guide or ESF
~Sstem Size Fluid ~in ~F FSAR Arrange-ment Inside/
Outside Primary Contain-ment Pipe Con-tainment to Outermost Isolation Valve T5I pe Test CI)
Potentral Bypass Leakage Path)>> Number TIRe Oper-ator Actuator Prima~r Mode secondarv Normal
<s)
Valvel+>
Shutdown Position Post-Accident Powe)Lm)
Failure Isola-tion closure Signal Time Power Source c()) No~es
"-83 Capped spare Z-85 Capped spare Z-86 capped spare Z-87 Capped spare .) 't Z 88A RHS safety 56 Yes Steam 12 6. 2-10 Outside 116'" See Note (Q3 valve discharge Sh. 14 to suppression pool za Z-88B RHR safety valve discharge 56 Yes Steam 12 6. 2-70 Sh. '4 Outside 106 '2 3/4" A No g1)
See Note Q3 to suppression pool Z-89A LMS from dry- 56 No Air 3/4 6. 2-70 Inside No ~ 2LMS4SOV152 Globe SOV Elec. Elec. Closed Closed Closed Closed B~ F,RM <1.5 Div II well Sh. 35 Outside 2I.NS*SOV153 Div I Z-898 Capped spare 3/4 Z-89C LMS from wet- 56 No Air 3/4 6. 2-70 Inside No q 2LMSI'SOV156 Globe scv Elec Elec. Closed Closed Closed Closed B,F,RM <1.5 Div II well Sh. 35 Outside 2LMS*SOV157 Div I Z-89D Capped spare 3/4 Z-90 ICS vacuum Yes Air 1 1/2 6. 2-70 Cutside 13' C
"'(n ) I 2 CS* NOV1 4 8 Globe NOV Elec. Manual Open Closed Open FAI FIRM 10 Div II breaker Sh. 36 Outside C 2ICS*NOV164 Globe NOV Elec. Manual Open Closed Open 'AI None 10 Div I Z-91A Instrument air 56 N 1 1/2 6. 2-70 Outside I Qll j IIX (a o) 2 IAS* SOV 1 67 Globe SCV Elec. Manual Open Open closed Closed Br FIRN <5.0 Div I to drywell Sh. 37 Inside 1 I 011 2IAS*SOV185 Globe SCV Elec. Open Open closed Closed B,F,RN <5.0 Div II'iv Z-91B Instrument air 56 1/2 6.2-70 Outside I Qll 2IASI'SOV168 Globe SCV Elec. Manual Open Open Closed Closed F,RM <5.0 I P
N BE to drywell 1
Sh. 37 Inside 1
1 I-Qll 2 I AS* SOV1 8 0 Globe SOV Elec. Open Open closed Closed BE F,RM <5-0 Div II Z-91C Capped spare 1 1/2 Z-91C Capped spare 1/2 1
14 ofW ApERTURE CARD
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~>O@y 'fa ago~ Nine Mile Point Unit 2 FS TABLE 6. 2-56 (Cont) location of valve Length of Inside/ Pipe Con- Valve( l3 Iso a-Pene-tration No.
~0DC System Reg.
Guide or ESF System Size Fluid ~in FSAR Arrange-ment Fixture<<r Outside Frimary Contain-ment tainment to cutermost Isolation Valve Type Test
<<a Potential Bypass Leakage Path<a> Number T~- --
Oper-t ~P Actuator Mode Seconder Normal
<ai Position Shutdown Post-Accident Power(t(F)
Failure Signal Closure Time
>r Power source cv>
Not Z-92 Sp~~ No Z-96 25~
Z-98A RHR relief valve discharge 56 Yes Water 3 6.2-70 Sh. 38 outside 207 '6>> 2CSL>>RV123 RV N/A N/A N/A N/A N/A None N/A N/A 2CSL+RV105 to suppression 2RHS*RV61A pool 2RHS>>RV110 2 RHS>> RV13 9 Z-98B RHR relief 56 Yes Hater 3 6.2 70 Outside 89'-8" 2C SHN RV1 1 4 RV N/A N/A N/A N/A N/A N/A None N/A N/A valve discharge Sh, 38 gz9) 2CSH>>RV113 to suppression 2RHS*RV61B pool 2RHS*RV61C 2RHS>>RV20B Z-99A Hydraulic unit 57 No Hy- 3/4 6. l-70 Outside <5'-0" No / 2RCS>>SOV68A Globe SOV Auto Remote OPen Closed Closed Closed B F RM <1 5 N/A, '2l from recirc flow draulic Sh, 39 manual control valve HYV 17A (drain line)
Z 99B Hydraulic unit to recirc flow 57 No Hy-draulic 3/4 6. 1-70 Outside <5 ~ -0>> N/A No r >>
i) 2RCS>>SOV67A Globe SOV Auto Remote manual Open Closed Closed Closed B~ F,RM <1 5 N/A,2<
Sh. 39 (3 control valve HYV 17A (open line)
Z-99C Hydraulic unit 57 No Hy- 3/4 6.2-70 Outside <5'-0' N/A No rQl~ 2RCS>>SOV66A Globe SOV Auto Remote Open Closed Closed Closed B F RM <1 5 N/A 26 to recirc flow draulic Sh. 39 manual control valve HYV 17A (pilot line)
APERTURE Ic CARD 15 of9
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,control 4 I 1)oeun1e> Nine Mile Point
~DOCKm'F'~ Vnit 2 FsAR l(EG~~ TABLE 6. 2-56 (Cont)
O~t'ize location of valve Length of Inside/ Pipe Con-Pene- FSAR Outside tainment to tration GDC or Arrange- Primary Potential Valve w 4t'so System Reg. ESF Outermost Type N . ment Contain- Bypass Guide stem Fluid ~in FiciureC 11 Isolation Test Leakage Position a-
~S ment Valve CC> Oper<< Actuator tion Z-99D Hydraulic unit P Number T~e ator Mode Normal Post Closure Power to recirc flow 57 No Hy- 3/4 6.2-79 Outside Secon~dar Cs) Shutdown Powey)E() signal Time Source:
draulic <5I 0n N/A Failure '
s a1 zl +:Notes-control valve Sh 2RCSc'SOV65A'- Globe--"-
5, N/A, C
39 SOV Auto Remote HYV 17A (closed Open Closed Closed Closed B ~ F, RM line) manual <1 r26 Z-100A Hydraulic unit from recirc flow control valve 57 No Hy- '/4 draulic
- 6. 2-7(
Sh.
Outside Dw No 2RCSSSOV68B Globe - SOV Auto HYV 178 (drain Remote Open Closed Closed manual ., Closed B ~ F~RM <1 5 N/A line) ( j26 Z-100B Hydraulic unit 3'.2-70 57 No Hy- 3/4 A to recirc flow Outside <51-0s control valve draulic Sh. N/A No HYV 178 line)
(open 39
(~A 2RCSSSOV67B Globe SOV Auto Remote manual Open Closed Closed closed BeFrRM <1 5 '/A, j; +26 Z-100C Hydraulic unit 57 No Hy- 3/4 to recirc flow 6. 2-70 Outside <5'-0" control valve draulic N/A No HYV 178 (pilot) line)
Sh. 39
(" 2RCS*SOV66B Globe .- SOV Auto Remote manual Open Closed Closed Closed B,F,RM ';<1.5,N/A '26 Z-100D Hydraulic unit 57 3/4 to recirc flow Hy 6. 2-70 Outside <5'-09 control valve draulic S)1 39 No , 2RCSSSOV65B Globe HYV 178 ('3)) SOV Autc Remote Open Closed (closed manual Closed Closed B F RM <1 5 line) N/A 26 All instrument R. G. Air/
lines reactor from 1.11 No Water 3/4 6.2-70 Sh. 41 Outside As close as vessel possible to EF check EFV N/A Auto valves N/A Open Open All instrument containment Open Open Excess N/A N/A : 27 lines penetra- R G. Air/ 3/4 6.2-70 Outside flow ting primary 1.11 Water Sh. 41 As close as "
possrble to (I() EFV N/A Auto N/A containment containment Open Open Open Open Excess N/A flow N/A , 27
~To be supplied in an amendment