ML17272A270
| ML17272A270 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 01/23/1979 |
| From: | Renberger D WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| GO2-79-18, NUDOCS 7901300123 | |
| Download: ML17272A270 (95) | |
Text
f~ = BFvULA'FARY INF()RMATI<)N DISTR IDUTI<)N SYSTEM
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ACCjPSION NBR -790130012 D()C.DATE: 79/01/23 N()WI'ZED~
NO D()CKET P FACIL>>50-397 NPPSS Nuclear Project-l, Washington Public Power: Supp 05000390
/AUTH.,NAME AUTH()R AFFI LIATI()N BENBEBGFB,D.L., 1'Iashington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION 0
VABGA,S.A. 'ight Hater Reactors Branch 4
SUBJECT:
Forwards responses to 781,103 First Bound conta'inment sys branch questions. Responses will be formally 'submitted as amend to FSAR.
0 DISTRIBUTION CODE:
BOO)B COPIES IIECEIVED-LTB J ENCL QO SIZE:
TITLE ~
PSAR/FSAR AMDTS AND-RELATED COBRESP()NDENCE.
NOTES:
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Washington Public Power Supply System A JOINT OPERATING AGENCY P. O. BOX 958 3000 GCO, WASHIHCTON WAY RICHLAND, WAIHINCTOH 99352 PHONC (509) 375 5000 Docket No. 50-397 January 23, 1979 G02-79-18 Director, Office of Nuclear Reactor Regulation U.
S. Nuclear Regulatory Commission Washington, D.C.
20555 Attention:
Subject:
Reference:
Hr. S.
A. Varga, Chief Branch No.
4 Division of Project Management WPPSS NUCLEAR PROJECT NO.
2 RESPONSES TO FIRST ROUND CONTAINMENT SYSTEMS BRANCH UESTIONS
- Letter, S.
A. Varga (NRC) to N. Strand (WPPSS), "First Round guestions on the WNP-2 OL Application - CSB", dated November 3, 1978.
Dear Hr. Varga:
Attached please find sixty (60) copies of responses to the referenced questions.
The few open items from the question set are being carried forward and will be submitted at the earliest possible date.
These responses will be formally submitted in the FSAR in an amendment within the next three months.
~ gQ'Pg Very truly yours,
~~yg R
,gg30 D. L.
RENBERGER Assistant Director Technology DLR:OKE:sg cc:
I. Littman, WPPSS, NY w/o JJ Verderber, B&R JJ Byrnes, B&R RC Root, B&R, Site HR Canter, B&R D. Roe, BPA FA Maclean, GE San Jose 790> 300 /~3 responses E.
- Chang, GE San Jose w/5 J.
Ellwanger, B&R NS Reynolds, Debevoise
& Liberman w/1 WNP-2 Files w/1 Q gV
4 1
d
"~
STATE OF WASHINGTON)
)
ss COUNTY OF BENTON
)
D. L.
RENBERGER, Being first duly sworn, deposes and says:
That he is the Assistant Director, Technology, for the WASHINGTON PUBLIC POWER SUPPLY SYSTEM, the applicant herein; that he is authorized to submit the fore-going on behalf of said applicant; that he has read the foregoing and knows the contents thereof; and believes the same to be true to the best of his knowledge.
DATED
, 1979 D. L.
RENBERGER
, 1979.
On this day personally appeared before me D. L.
RENBERGER to me known to be the individual who executed the foregoing instrument and acknowledged that he signed the same as his free act and deed for the uses and purposes therein mentioned.
GIVEN under my hand and seal this ~~ day of No ary Public in and for the State of Washington Residing at
'Pi>P-2 0 22.031 Provide a detailed ca"cula ion of the f iction loss coe ficient for the entire vent system.
Indicate whether the results of tne 4T (temporary tall test tank facility) portion of the ongoing generic Nark II test program have been used to confirm the calculation vent loss coefficient.
Additionally, indicate the margin applied to the calculated f iction loss coefficient to account for any differences between, the WNP-2 vent design and that of the 4T test facility.
Resnonse:
5'o known s"udies have been performed to exper'mentally determine 4T test downcomer vent loss coefficients.
However in their Pool Swell Analytical Nodel (PSAN)/4T test data comparisons (References 1 and 2), General Electric used downcomer vent
~
loss coef icients of 2.51 and 3.50 for the 4T test 20" down-comers and 24" downcomers, respectively.
These values were used as input to the General Electric PSAM and were calculated using information from Reference 3.
The QlP-2 downcomer friction loss coefficient (fL/D) that is used in pool swell studies is ecual to 1.9 (see Table 3.8-1).
Use of a value of 1.9 vs a
4T value ensures conse vatism
'n WNP-2 pool swell studies
'n that lower values of fL/D maximizes pool swell velocity (see Figure 4-24 of Reference
- 4).
References:
1)
NEOE-21544-P, Oec.
- 1976, "Nark II Pressure Suppression Containment System; An Analytical Model of Pool Swell Phenomena".
2)
Response
to NRC guestion 20.71 transmitted via letter HFH-275-78 to ter. J.
F. Stolz, Chief, Liqht liater Reactor Branch No. 1, USNRC, from Hr. L. J.
- Sobon, t1anager BHR Containment Licensing, General Electric Co.
on "Responses to NRC Request for Additional Infor-mation" (Round 3 guestions},
dated June 30, 1978.
3)
AEC-TR-6630, Handbook of H draulic Resistance-Coeffi-cients of Local Resistance and of Friction, I.E.
Idel chik, 1960.
4)
NED 21061 Rev. 3, tiark II Containment D namic Forcina Functions Information Re or t, June 1978.
'IHIP-2
- 22. 032 Provide the following information regarding the vacuum breaker systems between the wetwell and the drywell and between the reactor building and the wetwell:
a
~
Oescribe the preoperational and inservice tests that will be per-formed to verify that the setpoints of the vacuum breakers are at the appropriate pressure levels and meet the required opening times:
RESPONSE
b.
Preoperational and inservice testing of the vacuum breaker system is performed to verify that the valves open at the appropriate pressure levels.
For the single and double disk check valves, testing will be accomplished using a torque wrench applied to the disk pivot shaft and determining the opening pressure by correlation with the measured torque required to open the valve.
Correlation curves are provided by the valve manufacturer.
The response time for opening of the single and double disk check valves is not measured during preoperational and inservice testing.
Two of the double disk valves were tested by the manufacturer for compliance with the specification requirement that the valves be fully ooened within 1
second at 0.5 psi differential pressure.
Provide the sensitivity limits and hysteresis characteristics of the electrical switches.
Provide the results of your analyses of the maximum opening between the valve disc and the seat when the position indicator system indicates that the wetwell vacuum breaker valve is closed;
RESPONSE
The switches used for position indication of the wetwell-drywell and reactor building-wetwell vacuum breakers are'a contact-probe type.
These contact probes are very sensitive and have zero hysteresis.
Accuracy within 0.010" is possible.
Valve closed indication is taken directly from the valve face.
Four probes are located 90 apart, straddling the valve vertical centerline.
Oue to the accuracy of the switches (0.010"), the location of the four probes, and based on the geometry of the vacuum breaker, the maximum opening between the valve disc and the seal when the position indicator system indicates a closed valve is 0.012".
WNP-2 c.
Provide a schematic of the vacuum breaker assembly.
Provide your analysis of the minimum flow area and the total loss coefficient for one vacuum breaker assembly.
RESPONSE
The attached figures (2-1 and 2-2) from the manufacturer's instruc-tion manual illustrate the configuration of the double disk vacuum breaker
- assembly, and the seal detail.
FSAR Reference 3.8-8, which was provided to the NRC, also provides illustrations and details of the vacuum breaker system and the Anderson, Greenwood 8
Co.
vacuum breaker valves.
The minimum flow area of the vacuum breaker valves is 295.6 square inches based on the 19.4 inch diameter inlet orifice.
Capacity certification tests were performed on 3 wetwell-drywell and 3 reactor building-wetwell vacuum breaker valves.
Using the flow data from these tests, resistance coefficients (K) were determined as follows:
l<etwell-Drykell valves - 4.73 at 0.360 psid Reactor Building-lletwell valves - 1.65 at 0.289 psid These values 'of K include valve entrance effects and are based on the connecting piping internal area of 424.6 square inches.
I 4
0
~,ANDERSON, GREEN D 6 CO.
(
REPORT NUhQKR 05-9040-016 Page 4 LIFTING EYE BOLTS
~ 0 Pivot Arm Shank F
~
Turnb ckle Shank rm Shank Arm Turnbuckle Pivot Arm Disc Shank Disk FIGURE 2-1.
VALVECROSS SECTION (DUALDISC)
FORM SRd
0 ly e
t
O l
AHEERSOH, SREEHWtl 6 CO.
1 IUEPORT NUhQSR 05-9040-016 Page 5 2-4.
VALVE SEAL. A circular seal around the perimeter of the valve orifice provides a superior seal when the valve is in the closed position.
The details of this seal are shown in Figure 2-2.
Use the small sketch labeled COMPLETE VALVEfor orientation, then refer to the CVl-Lseal detail.
The valve disk assembly (cross hatched) is the movable-
, part. All other ports are fixed.
A circular moat encircles the valve orifice.
Over this moat the valve seal diaphragm material is stretched and held in place by two concentric retainer rings.
The valve seals when the seal lip on the volve disk assembly touches this diaphragm.
The inner of the two seal retainer rings serves to limit the penetration of the lip into the seal from 0.010 to about 0.020 inches ond also serves as a secondory metal-to-metal seal.
The primary seal between the lip and diaphragm is pressure boosted.
Pressure From the high pressure, or downstream side'is Fed into the moat under the seal diaphragm.
This causes a zero pressure difference across the seal at point A and a full pressure difference across the seal diaphragm at point B.
COMPLETE VALVE BOOSTER PASSAGE~
p l
I I i lg I
i
~
SEAL MATERIAL MOAT SEAL RETAINER SEAL RETAINER PRIMARY SEAL VALVEDISK ASSEMBLY SECONDARY METAL.TOMETALSEAL IMBEDDED MAGNETS FlGURE 2-2.
CVl-LVALVE, SEAL DETAIL
p'i 022.033 You state in Section
- 6. 2. l. l. 8. 2 of the FSAR that operation of the containment purge system will be limited to one percent of the reactor operating time.
Me find this'pproach acceptable provided you:
a ~
Expand your definition of reactor operating time to include the three operational modes of startup, hot standby, and hot shutdown.
b.
Demonstrate that the purge system isolation valves can be closed when subjected to the.environmental conditions, including pressure, that occur following a postulated loss-of-coolant accident.
c.
Combine the time to purge the suppression chamber with the time to purge the drywell in the proposed one percent restriction on the operating time of the containment purge system.
~Res ense:
a
~
Refer to response to question 022.019.
b.
Refer to response to question 022.019.
C.
Purge system operation, addressed in paragraph 6.2.1.1.8.2, includes purging the drywell and suppression chamber.
0 WNP-2 22.034 Identify all access openings to the secondary containment and discuss the administrative controls that will be exercised over 'them.'iscuss the instrumentation to be provided to monitor the status of the openinqs.
Indicate whether position indicators will provide readout information to the plant operator and whether alarms will be annunciated in the main control room.
RESPONSE
All access openings to the secondary containment are administratively controlled.
Details related to this administration and the indications provided are presented in the WNP-2 Security Plan, transmitted under separate cover to the NRC.
Q 22 0
ov',ie ho 'o owing information related to potent'al bvoass leakage Oaths:
For each air or'water seal, perform an a.-.alysis of tne.fluid inventory which wi'1 bo ava'lable to maintain, the seal for 30 da:s fol';'ii.c a
l~~
pos ulated loss of coolant accident and de.-,.on-strate that this fluid inventorv will be C 'u"ficient.
Describe the testing program and "he specific details of your proposed techn'cal specifications which will verifv the ass'..ptions used in the analysis.
Provide the bas's fo" the valve fluid leakage used in your analys's.
For each of these paths where water seals e1imina te the po ten tia1 or. bypass
- leakage, provide a sketch showing the loca t'n o= the water seals relative to the sys"em isolation valves'.
E:)plain why the combustible gas contro'ystem is omitted from Table 6.2-13 o
the FS)NR as a
po"ent'al leakage path.
Der;ons" rate ti a-system meets each of the pi.ovisions of Branch Technical Position CSB 6-3, Sec" on B-9,. fo a
c osed system
~
Respo::se 4
~
Potential bypass leakage path" round "he secondary containment are discussed in 6.2.3, Secondary Containment Functional Design.
As discussed in 6.2.3.2, the two 24 inch reactor feedwater (AF<<l) lines are the only lines o
which a water or air seal is assumed wh'ch p"events secondary containment bypass leakage.
.:n analysis was performed which showed th. t a vertical water seal greater tha".. thirtv feet will remain between the isolation valves and the containmcnt atmosphere.
As discussed in G.2.3.3 and G.2.6.3, the isolat'on valves on the lines which were ideiitificd as poten'ia'ypass leakage paths around the seconca v con.
tainment, will be tested to ensure "hat the
zn<<i~vidual leai.~go rates a
e below tno 1'~x"s set by ASl.:E Co<<ie Section '.I, Tab'e il743~~2v-l.
The limit a 1 lowec by Table itF3 ~i 20-1 was "he value assume<<'.
in calculating the water lost from the water seal through the B."-~~'scla"'on.
valves.
-igure 6.2-25 s'. ows the
!3FY iine routinr.
The water seal will..ot drop belo~; the horizontal run o piping on the 543 foot elevat'n.
c.
The containment atmosphere control (CAC) system is a closed system outside the primary contain-ment.
Suction and discharge are to the pr'mary containment.
All piping remains with n the seconda
) containment o
Any lear ago K
om the Cl~'C system will be processed bv the standby gas treatment sys"em prior to release to the environment.
The CAC system is described in detail in 6.2.5 and shown in E'igure 3.2-17.
The CAC system meets all the criteria stipulated in BTP CSB 6-3 para<<;raph P.9.
The CZ~C system docs not directly communicate with the envi on-ment, is designed to Code Group D stan<<',ards, meets Seismic I design requirements,
" s cosigned
=o " e primary containment pressure an<<'. temper-ature design conditions, 'is <<lcsigned
- against, the consequences of any breach in the reactor coolant prcssure boundary (pipe whip etc.)
d wi) 1 I
g an 1 be op<<.n to the primary conta'nment atmo-sphere during the intergratcd leak rata test.
Xn addition, the CAC system can be isolated from the primary conta'nment by two redurdan" 's a ion valves.
There is no rea on to consice the CAC system as a secondary containment bypass leakage path.
WNP-2 022; 036 It is our position that safety-related equipment located in the drywe]l should be exposed for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to a saturated steam environment at'340 F
and a pressure equal to the drywell design pressure during the qualifi-cation testing.
(Refer to Item 022.021 transmitted on September 18, 1978).
RESPONSE
Safety related equipment in the drywell for WNP-2 was procured before 1974 and was specified and qualified to a saturate]
steam environment for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> at 340 F and 6 additional hours at 320 F at 45 psig (drywell design pressure).
These environmental conditions formed the design basis for WNP-2 in the drywell and were consistent with IEEE standard 323-1974 (Table A2), "Test Conditions for Boiling Water Reactors" when it was published.
0
0 WNP-2 22,037 Discuss in detail the design provisions incorporated for periodic inspection and operability testing of the individual components of the containment heat removal
- systems, including the pumps, valves, duct pressure-relieving devices and spray nozzles.
RESPONSE
The containment heat removal system is an operating mode of the residual heat removal (RHR) system shown in Figure 3.2-6.
Operation of the RHR system for containment heat removal is discussed in 6.2.2.
All power operated valves can be exercised without affecting reactor oper-ation.
All RHR check valves in the primary containment, RHR-V-41A, B and C and RHR-V-50A and B, can be remotely exercised by means of a pneumatic actuator (see Note 3 of Table 6.2-16).*
The check valve on the discharge of each RHR pump will be exercised when the RHR pump is operated.
All relief valves can be removed and bench tested when the RHR system is not required.
The RHR pumps have a full flow test line back to the suppression pool which allows testing of the pumps without affecting reactor operation.
Sufficient instrumentation is available to verify proper flow, NPSH, and discharge pressure.
The pumps are in an accessible area outside the primary containment where, if necessary, they can be locally monitored.
The spray nozzles on the containment spray headers are passive components.
Each drywell spray header contains 150 spray nozzles.
Each spray nozzle consists of a nozzle body and seven removable spray caps.
Each spray cap has an internal vane.
Since the spray nozzles are passive components with no moving parts, testing is limited to a qualitive air flow test during preoperational testing.
The RHR heat exchangers are periodically used for cooling down the reactor pressure vessel after-shutdown in preparation for maintenance and refueling.
This verifies operability of the RHR heat exchangers.
There are no duct pressure relieving devices, or anything similar, on the RHR system.
In addition, the RHR system is built to Code Group B (AStlE III-2) standards and is subject to the applicable inservice inspections discussed in 6.6.
- A revised Table 6.2-16 is submitted with question 22.044.
- 22. 038 Provide a detailed analysis of the available net positive suction head for the pumps in the reactor heat removal systems that are used as part of the containment heat removal system.
This analysis should demonstrate compliance with the guidance contained in Regulatory Guide 1.1, "NPSH for Emergency Core Cooling and Containment Heat Removal System Pumps."
Indicate the required net positive suction head.
RESPONSE
The net positive suction head (NPSH) for all emergency core cooling system (ECCS) pumps was calculated in accordance with Regulatory Guide 1.1.
= Wetwell air space pressure
+ static pressure-friction losses
- vapor pressure Static head equals minimum suppression pool water level, 466 feet, minus centerline of RHR pump suction nozzles, 421 feet.
Static head equals 45 feet.
Friction losses for suction piping is approximately 3 feet for all the RHR pumps.
The suction strainer is assumed to be 504 plugged.
Vapor pressure at the peak suppression pool temperature of 220 F is 2.5 psig (6 feet).
In accordance with Regulatory Guide 1.1, "no increase in containment pressure from that present prior to postulated loss-of-coolant accidents" is assumed.
Therefore, the wetwell air space pressure is assumed to be 0 psig, even though maximum suppression pool temperature is 220 F.
This is conservative but not realistic since the suppression pool will bg at saturation pressure any time the suppression pool water'exceeds 212 F.
Hased on the above, the NPSH available is 36 feet.*
The NPSH required by the pump manufacturers as documented by pump performance curves is ll feet at 7450 gpm rated flow.
See figures 6.3-10a, b,
c for pump per-formance curves.
- Page 6.2-19 will be revised per the attached draft.
4%l
NNP-2 for accident protection including support structures are designed in accordance with Seismic Category I criteria (see Chapter 3). The available NPSH was calculate6 i accordance with Regulatory Guide 1.1, and is 4.
The pump characteristics for the LPCI pumps are shown in Figures 6.3-10a, b,
and c.
The LPCI system incorporates a relief valve on each of the pump discharge lines which protects the components and pip'ng from inadvertent overpressure conditions.
These valves are set to relieve pressure at 500 psig with a capacity of 25 gpm.
The common suction relief valve on loop."A" and "B" is set at 220 psig with a capacity of 25 gpm.
The suction valve on loop "C" is set at 125 psig with a capacity of 10 gpm.
Provisions are included in the LPCI system to permit testing of the system.
These provisions are; a.
All active LPCI components are designed to be testable during normal plant operation and/or during plant shutdown as discussed in 6.3.1.1.2m.
b.
A discharge test line is provided for the three pumps to route suppression pool water back to the suppression pool without entering the re-actor pressure vessel.
c.
A suction test line, supplying reactor grade water, is provided to test loop "C" discharge into the reactor pressure vessel during normal plant shutdown.
d.
Instrumentation is provided to indicate system performance during normal and test operations.
e.
All check valves and motor-operated valves are capable of, operation for test purposes.
f.
Shutdown lines taking suction from the recir-culation system are provided for loops "A" and "B" to provide for shutdown cooling and to test pump discharge into the reactor pressure vessel after normal plant shutdown.
g.
All relief valves are removable for bench-testing during plant shutdown.
6.3-19
WNP-2 g
22.039 Describe the analysis performed to establish the size of the suction screens in the reactor heat removal system.
Provide a drawing showing the suction screen assembly.
RESPONSE
The screen size for the. suction strainers on the residual heat removal system is based on the more restrictive of the criteria set by the pump manufacturer or the spray nozzle orifice opening.
The pump manufacturer imposed a maximum particle size of three thirty-seconds (3/32) of an inch based on the size of the smallest orifice/
flow path in the pump mechanical seal.
This is significantly more restrictive than the, requirement imposed by the spray nozzles which have an ori-fice opening of seventeen sixty-fourths (17/64) of an inch.
Accordingly, the strainers will be specified to prevent the passage of particles three thrity-seconds of an inch or greater.
The suction strainers are presently being procured.
A drawinn will be supplied at a later date.
0 WNP-2 g 22.040 Provide a full scale drawing for Figures 3.2-2, 3.2-3, and 3.2-6 of the FSAR.
RESPONSE
Enclosed are seven copies each of the requested full scale drawings, (mailed separately)
- 22. 041 Table 6.2-16 of the FSAR does not identify the specific criterion of the General Design Criteria that applies to the isolation pro-visions for a number of systems.
Accordingly, provide the specific criterion for the following systems.
RCIC pump minimum flow bypass (X-65) RCIC turbine exhaust to suppression pool (X-4), RCIC turbine exhaust vacuum breaker line (X-116),
RCIC vacuum pump discharge to suppression pool (X-64), RCIC pump suction from suppression pool (X-33),
LPCS line (X-63),
HPCS pump suction (X-30),
LPCS pump suction (X-34),
HPCS line (X-49), drywell spray (X-llA, B), sup-pression pool spray, (X-25, A, B),
RHR lines (X-47, 26, 117, 35, 32, 36, 116, 48, RFW to reactor (17-A, B), Suppression pool clean up (X-100, 101, drywell equipment and floor drain (X-23, 24), contain-ment ventilation ((X-53), 66,
- 119, 3, 96, 99, 105, 98,
- 103, 104, 97,
- 108, 67.
Refer back to item 022.014 transmitted to you in our letter dated September 18, 1978, for our position on Note 2 of Table 6.3-16.
RESPONSE
The General Design Criteria for all penetrations including the above are included in revised Table 6.2-16 submitted in response to question 22.027.~
022.042 Table 6.2-16 indicates that the isolation provisions for the recirculation pump seal (x-43, A, B), CIA for HSIY and HS relief valves (x-56), CIA and nitrogen backup to the ADS valves (x-89, A, B, 91) conform to the require-ments of Criterion 57 of the General Design Criterion (GDC).
It is our position that the isolation provisions for these specific lines should meet the requirements of Criterion 56.
- However, a single isolation valve outside containment is acceptable as discussed in Section 6.2.4 (II.3e) of the Standard Review Plan (SRP).
Revise Table 6.2-12 to reflect our position and indicate if the other acceptable alternatives for meeting the requirements of the GDC as specified in the SRP could be applied to
'any of these lines.
RESPONSE
The General Design Criteria.for the above penetrations has been changed from 57 to 56.
Table 6.2-16 has been revised to indicate this (see the response to question 22.027).
Containment isolation is provided by a check valve and a motor operated gate valve (see also the response to question 22.043 for the justification for locating the check valve outside containment),
Tab's
- 6. 2-',
- 6. 2-16, and 7. 3-13 of the FSiXR indicate ti'.at a c!'cck va've outside the containment is considered as a
contain.-., nt.'soiation valve for the minimum flow at the pumps the reactor heat removal system
(::-47, 48),
vacuum re3.ief from secondary containment
(~;-66, 67, 119) and a process samp e 1'ne (Y-69D).
Pr'ovide justification 4or this design approacn.
Response
Tables 6.2-13, and 7.3-13 have been deleted.
See question 22.027 'r revised Table 6.2-16.
There are check valves inboard of the isolation va3.ves on the minim m flow line from the RilB. pumps (Y,-47, X-48).
These valves are built: to the same standards as the isolaton valves i'1 and wi ', i" necessary, isolate the minimum flow line from
":".e p imary containment;
- however, these chock valves are not considered containment isolation valves.
Please see rev'sed Table
- 6. 2-3.6.
There a"e no check valves on the process sample line
(>:-69D).
Ti.e rota"ion, C.V., which was previously uqcd was not intended
-o cesignaie cneck valve.
Flevisod Table 6.2-16 now clea y
designates ti e valve tvpes for the isolat on valves on pene'ration Y-69D.
Doth 'so'a"ion valves on the reactor build'ng to wetweli vacuum el'ef 1'-nes (Y.-66, X-67, and Y-119) are located out-side le wet~;e11 to improve valve operabili ty (see Note 17 o" re'ised Tab'e 6.2-16).
The reactor building to wetwell vacuum re'ef system is required to prevent excessive negative pressures
'in the primary containment under certain postulated cord'ions (see 6. 2. 1. 1.4).
Thc disc in the check va ve is main"'a'ned in the close position during normal ope"at'on by means of a spring actuated lever arm and magnets embedded in the periphery of the d'sc.
The magnet'c and spring fo ces are overcome, and the d'sc starts. to open, when ti.e pressure differential across the valve exceeds 0.2 psid.
The check valves have position indication 1'ghts which can aier" the operators to the fact that a chock valve
's not uliy closed.
The operator can t!ion remotely shut v
ve means of a pneumatic operator.
Tne operat'ng switch 's spring-return to neutral.
The ai" supply to "nese va'es Quali" s Class T..
<:!:P-2 i?evised Table 6.2-16.
now lists check valves outside containment
'.or C:i'. to "he inboard YiSiV's and i'IS relief valves (X-56) and C:?: and noc~en backup to the PDS valves (2-89K and B).
These check valves are in all three cases inboard of motor Gpe avcd isolation c;lobe valves.
The check valves are located outside of the primarv containment to improve valve ope ability as discussed in Note 17 of Table 6.2-16.
Q 22.44 Revise Table 6.2-'16 of the FSAR to identify the leakage detection provisions for those systems that rely on remote-.
manual isolation.
This revision should demonstrate con-formance with the guidance contained in Standard Review Plan 6.2.4.
Containment Isolation Systems, which states that provisions should be made to allow the operator in the main control room to-know when to isolate systems that require remote manual isolation.
While you have responded to a similar question in Amendment No.
1 to the FSAR (Item 022. Olid), Table 6. 2-16 should nevertheless be expanded.
R~es onse:
See revised Table 6.2-16.
Reference is made in the ""Isolation Signal" column to the note in the table which discusses isolation signals generated by the individual system process control signals or remote-manual closure based on information available to the operator in the main control room,
0 e
'vI
TABLE 6.2-16 PRI NARY CONTAINHENT ISOIATION 0
a
>>4 4>>o I.INE DESCRIPTION a0 4>>
g O~
>>>> 4>>>>
O >>ll Rg a
~~ 34 B.
xi n>>
>>a o>>
04 8
rl o
>>4>>
a o o gg 0
4>>
I>>
4>> Ul O
o o
B o 4 ao 4>> 44>>
44>>
0 0
HS Line h ISA 3 ~ 2 2 3.2-25 6.2-31$
HS-V-67h
)IO Cate
}lSLC-V-3A'IO Cate 0
Ac 0
Ac 55 h
HS-V-22A AO I hir Globo HS"V-28h AO 0 Air Globe hir/
Spr hir/
=Spr Ac Ac B>>c>>
C>>D, P,H B,C, G>>D>>
P>>H B>>c>>
C,D, P,ll 30 00/CC C
26 3 No S
Valves T.B:
No I ~
15 0
0/c c
c 26 3-10 4
No S
Valv<<s T.8, Ho 1, 15 C
C 0
AS-IS 1-Std 10 Y<<.s S
Valves R.B.
1/2 0
C C
AS-IS 1-Std 5
No S
Valves T.B.
No 15 NS I.lno 8
)LS l.ln<<. 0 HS-V-678
}IO Cato HSLC-V-38 }IO Cato 0
Ac 0
Ac 18C 3.2-2 55 h
}IS-V-22C AO I hir 3.2-25 Clobu 6.2-31$
HS-V-28C AO 0 hir Clobe HS"V-67C
}IO Cats
}5LC-V-3C HO Cate 0
AC 0
Ac 188 3.2-2, 55 h
HS-V-228 AO I hir 3.2-25 Globo 6.2"31$
}IS-V-288 AO 0 hir Globo hir/
Spr hir/
Spr Ac Ac hir/
Spr hir/
Spr AC AC B,C, C,D, P,H B,C, G>>D>>
P, ll B>>C.
C,D, P,H 30 B,c, C,D, P>>H B,C, G,D, P,H B,C, C,D, P,H 30 0
0/c c
c 26 3 No S
Valves T.B.
No 1, 15 0
0/C C
C 26 3-10 4
No S
Valves T.B.
No 1 ~
15 0
0 0
AS-IS 1-Std 5
1/2 No S
Valves T.D.
No 15 C
C 0
AS-IS 1-Std 10 Yus S
Valves R.B.
I/2 No 0
0/C C
C 26 3 No S
Valves T.B.
No 1, I5 0
0/C C
C 26 3-10 6
No S
Valves T.B, Nil I ~
15 0
0 0
ASIS 1-Std 5
1/2 Ho S
Valves T.B, Ho 15 C
C 0
AS-IS 1-Std 10 Yes S
Valv<
~'
4 ve>>4 4>>
>>e 'r>>>4>>
4 s'>>4>>>>
~ se wrqq~
TASLE 6.2-16 (Coatlaucd)
LINK DESCRIPTIOH HS Llue D
~>
O a
4 a
a a
o j
o>>>
o
'>>4 Hs-V-67D H0 Cata HSLC-V-3D H0 Cata 0
Ac 1SD 3.2-2 SS h
HS-V-22D AO I hlr 3.2-2S Globo 6 ~ 2-31)
HS-V-28D AO 0 hlr Cloba oR
>>>>4 0
~>
hir/
Spr Air/
Spr AC hc S,C, SH C>D>
P,H S,C, RH C,D>
P,H S>C, RH C,D, P,H SH
~4 g I>>>
A
~4 Eo o
>4 a
4>
Q 0 g o1f 0
>4
>>>c
>4 l>
B 39i 8
a>>
3A
>c 40 4
~4 Cf
>>>4>
o
>> f
>4 fe4 Osdh X 00/C C
C 0
0/C C
C 0
C C
AS"IS 1-Std I/2 C
C 0
AS-IS 1-Std 1/2 S
Ho S
Valves T.BE No 1S 10 Yee S
Valvea R.S.
No 26 3 No S
Valvaa T.H.
No 1 ~
1S 26 3"10 6
No S
Valvea T.S.
No I, 1S HS Llua Drala 22 3>2-2 SS h
HS-V-16 6>2-31!
HS-V-19 H0 Cato H0 Cata I
AC Ac B>C>
SH C,D, P,H a,c, C>D>
P>H 0
C C
AS-IS 3
Std No S
Valvca T.B.
0 C
C AS-IS 3
Std 6
No S
~ 19
~
~
~
TABLE 6.2-16 (Continued) 5 ~
SRAPT UNTIL MPPSS RRVIRM
\\
LlHR l)ESCR1PT1ON RPM Lfno h 17h 3.2-2 55 h
RPM-V-10h 6.2-3lb RPM-V-.32A RPM-V-65h RMCO-V-40 Chock PC Check NO Cate
}Q Gate 0
AC 0
0 Process Pro/Spr 0
0/C 0
0/C 31 Nanual 0
0/C
~'/5 0
0 0/C -
24 0/C -
24 0/C AS-IS 24 C
AS-IS 6
Ho 2
Ho 8
Ho M
24 No M
M Valves T.B. l.5 l6 M
RPM Line B 178
- 3. 2-2 55 A
RPM-V-10B 6.2-3lb RPM-V-32B RPM-V-65S RMCU-V-40 Check PC Chock NO Gate HO Gate z
4 Process 0
Prccosa 0
AC 0
AC Process Pro/
Spr AC 0
0/C 0/C -
24 0
0/C 0/C 24 No 2
Ho M
Valves T.B. 1.5 16 M
31 llanusl 0
0/C 0/C AS-IS 24 &8 8
Ho M
0 0
C AS-IS 6
5l8 24 No M
Cylinder Shut tlo Bruin 76b 76e 76c HY"V-18h HY-V"19h HY-V-20h Cylinder Cylluder Shuttle Bra)n 77f 77b 770 77c HY-V-17B ItY"V-IBB BY"Y"19B HY-V-2OB RRC Hydraulic Lines 3.2-3 57 8
Cylinder 76E HY-V-17A SO Globe SO Globe SO C)obo SO Globo SO Globe SO Globe SO Clobe SO Globe 0
AC 0
AC 0
AC 0
AC 0
AC 0
AC 0
AC 0
AC Spring S,F RH Spring B,P RN Spring B,F RH Spring S,F RM Spring S,F RH Spring S,P Lt Spring S P RN Spring S,F RH 0
0 C
C 3/4 <5 5
Ho ll Valves RB Hu 28 0
0 C
C 3/4 <5 5
0 0
C C
1/2 <5 5
0 0
C C
1/2 <5 5
0 0
C C
3/4 <5 5
Hu H
Valves RB Ho 28 0
0 C
C 3/4 <5 5
0 0
C C
1/2 <5 5
0 0
C C
1/2 <5 5
~l
~
~ e ~ e Kr
~ rr YABLE 6.2-16 (Continued) ry
~ g r ~
DRAPT UNYIL MPPSS REVIEH I
LINE DESCRIPTION HPCS to Reactor 6
))pcs-V"4 )0 0
'ate hc 46 Hanual 3.2-7 SS A
HPCS-V-S Chock I Process Process 6.2-31L c
c o/c-C C
O/C AS-IS 12 12 17 9
Ycs I Valves R,bi Ho 3, 24 LPCS to Reactor 8
3i2 7 A
LPCS-V"6 Chock I Process Process 6r2-31L c
c 0/c-12 Yes N
Valves R.b.
No 3, 24 HPCS poop suction 31 fron suppression pool Ll'CS-V-S HO 0
AC Gate Srg-7 56 b
HPCS-V-lS )0 0
AC 6,2-31n frrtrf AC 46 Hanual C
C 0/C AS-IS 12 27 22 AC 46 Hanual C
C 0/C AS-IS 18 18 3
Yes ll Valves R.b.
Ho 18, 2.'/
LPCS puup suction 34 3o2-1 56 b
LPCS-V-l )0 0
Ac 6r2-31n Cote AC 46 Hanuai 0
0 0/C AS-IS 24 Std 2
Yes ll Valves R.b.
No Igi2 Q HPCS test line 49 NPCS puap nfn. flou HPCS suction relief NPCS disc)urge relief LPCS test lfno 63 LPCS puop nfn.
f lou LPCS suction rclfcf LPCS dfscharge relief 3r2 7 56 b
NPCS-V-23 6.2-3lf lfPCS-V-12 PCS-RV-14 IIPCS-RV-35 3r2-7 56 b
)PCS-V 12 6r2-31f Ll'CS-'V"11 LPCS-RV 31 LPCS-RV" 18
)0 0
Globo
)0 0
Cato Relief 0 Ac Ac
)0 0
Clobo
)0 0
Globe Relfof 0 Ac AC PP Relief 0 PP Relief 0 PP AC Spring Spring AC Ac Spring SPrfn8 FgX3 RH 38 RH P,X3 RH 38 RH C
C C
AS-IS 12 Std 6
- Yes
'M Valves R.B, No 18 C
C C
AS IS 4
4 S3 C
C C
C C
C 1
2 65 10 19 19 C
C C
C 0
0 50 19 19 C
C C
AS-IS 12 Std 4
Ycs M
Valves R.b.
Ho 18 C
C 0/C AS-IS 3
Std 81 SLC to Reactor 13 3.2-S SS h
SLG-V-7 6o2-31n Check I Process Process SLC-V"4A SLC-V-48 Explo- 0 siva Explo- 0 siva AC AC SLC-V-6 Check 0
Process Process C
C C
C C
C C.C C
C C
C 1-I/2 1
1/2 1-1/2 1-1/2 136 136 No Q
Valves R.b, No 21 21 P
'0 I
~ I
TABI.E 6.2-16 (Continued)
~
'RAPT HHTIL MPPSS REVIEM LlNE DESCRll'TIOH D)l S<<tv lee I.lna 92 RHR Cu>>dousing 21
}Iud>> SL>>elm Supply 9.2-4 56 B
6.2-311.
3.2-8 55 h
6.2-3le DM-V-157 DQ-V-156 RCIC-V" 63 RCIC-V-76 RCIC-V-64 Gate Cate Ho Cate HO Globe
}$
Cate I
Hanual 0
}b)nual I
AC I
AC 0
}fanual Hanual AC K
2 RH 0 %
0/0 AS-IS 10 16 0
0 0
As Is 1
5 C %
C.
AS-IS 10 16 Ho ll Valves S.B.
~ 13 Yuu S
Valve>>
R. 8, Ho RC1C T>>rblnu Stcam 45 Supply RCI C l Ump Hlnllxuxl 65 Pluu 3.2-8 55 A
6.2"31e RCIC-V" 63 RCIC-V-76 RCIC-V 8 3.2-S 56 B
RCIC V-6.2-314 19
}$
Cato
}lo Clobe
}$
Cato AC I
AC 0
}$
RH K
RH g'~
RH 32!
33 0 %
0/C AS-IS C
C C
AS-IS O %
O/C AS-IS 4
Sca 0
0
'C As Is 2
5 7
Ho M
Valves R.B.
Nu 22 10 16 Ho S
Valves R.B ~
Hu 1
5 RCIC T>>rbluu 4
Exbuusc RCIC Turb}no 116 Exbaust Vucu)!m Br<<aber 3.2-8 56 B
6.2-3ln 3.2-8 56 B
6.2-311 RCIC V 68 RCIC-V-110 RC'10-V-113
}$
Cate HO Gate HO Cato DC DC 32f 35 Xl RH Xl RH 0
0 0/C AS-IS 2
Std 9
Ho h
Vulveu R.B.
Ho }7 O
O O/C AS-IS 2
Sca 5
Hunual 0
0 0/C AS-IS 10 Std 10 No S
Vulveu R.B.
Hu 22 RClC Vuc>>um Pump 64 Dlucbargu 3.2-8 56. B RCIC"V" 6.2-3lq 69 HO Cata 0
DC DC 36 Hanual 0
0 0/C AS IS 1-Scd 1/2 4
Ho 9
Valves R.B.
Ho 22 ECIC Pump Suction 33 from Suppress!on Pool 3.2-8 56 B
RCIC-V-6.2-31n 31
}Io Cato 0
DC DC 32 Hanual C
C 0/C AS-IS 8
Scd 2
No N
Vu)ves R.D.
No 23 Rl'V II<<aa Spray 2
- 3. 2-8 55 h
6.2-31o RCIC-V-66 RCHr V 13 RIIR-V-23 Ho Gate
}$
Globe 0
DC DC Check I Process Process 3/j 34 A,U, H,X2 RH c
o o/c 6
Ha M
Valvus R.B.
Ho 3 C
0/C 0/C AS-IS 6
15 2
No M
Valves R.B.
No C
0/C C
AS-IS 6
Std 7
Yeu II Valves R.B.
No
~o a~g 4' r
ig
~ * 'ee
~.y> ~
~
~~,
+s tycho i i i v
~
se ~'a 'v'g a~.~a appal;1 ipse~
~
v YABLR 6.2-16 (Contjunco)
I 1
~ nl DRAPT U)fIILUPNESS REVIGf LINK DESCRIPTION Dryucll Spray loop A llh 3o2-6 S6 B
I'LL-V"
- 6. 2-3)B 16A RNR-V 17A HO 0
Cato
)Q 0
Cato AC 46 Hanual C
C 0/C AS IS 16 10 26 Yea M
Valves R.Bi AC 46 Hanual C
C 0/C AS-IS 16 10 24 No 17, 24 Dryuel1 Spray Loop B 1 1B
- 3. 2-6 56 B
RBR-V-6.2-318 16B RNR-V-17B
)IO 0
Cato
)LO 0
Cato AC AC 46 Nanual C
C 0/C AS-IS 16 10 12 Yea If Valves R.B.
No 17, 24 AC AC
'6
)faunal C
C 0/C AS-IS 16 10 2
LPCI Loop h LPCI Loop B LPCI Loop C 12A 3o2-6 SS h
RBR-V-6.2-3)L 41A RIIR-V 42A 12B 3.2-6 SS A
RIR-V-6.2-31L 41B RNR-V-42B 12C 3.2-6 SS h
MR-V-6,2-31L 4)C RBR-V 42C Check I Process Process
))0 0
AC Cata AC Chock I Process Process HO 0
))O 0
AC Cato Check I Process Process C
C 0/C 14 46 Hanual C
C 0/C AS-IS 14 12 20 C
C 0/C 14 46 Hanuel C
C 0/C AS-IS 14 12 20 C
C 0/C-46 Hanuel C
C 0/C AS-IS 14 12 21 Yes M
Valves R.B.
No 3, 24 Yca If Valves R.B.
No 3, 24
'Yos
)I Valves R.B.
No 3, 24 Shutdovn Cooljnk Return Loop h 19h 3>2-6 5S A
RBR-V-6+2-31ja 504 RNR-V 123A RBR-V 53'O I
AC Cato
)IO 0
AC C)oba AC AC Check I Process Process P,h,
)Dj V>H, R2
- Hph, R)I Ut22 C
0 0
12 0
0/C C
AS-IS 1
C 0
C AS-1$
12 40 5
Yes U
Valves R,'B ~
No 3 Shutdoun CooljnS Return Loop B 19B
- 3. 2-6 55 A
RIBL-V 6.2-3ha SDB RBR-V-1239 RBR-V-53B HO I
AC Cato YO 0
AC Cloba AC Cllcck I Process Process F)A,
)Of ll,H, 22
)Iih, RH U,I2 C
C 0/C 0
AS-IS 1
C 0
C AS-IS 12 40 2
12 'es lf Valves
- ReB, No 3 Shutao~ Cooljnk Suctjoo 20 3o 2-6 5$
h IUE-Y-9 6o2-31k RBR-V-S HO Caco
)LO 0
Cate A,U, IOI
)f,)I2 hyVy F)j Hpk2 C
0 C
AS-IS C
0 C
AS-IS 20 40 Yca M Valves R.B, No 20 40 14
J r
~
~
~ 8>
> I
- I
\\
~
h 4
> ~ 1
~ H4 '>~~ >>>A YABLE 6.2-16 (Continued)
DRAFl UNTIL ttPPS5 RKVIEtj LlNE DESCRtPTIDH Rtlb Loop At puap test line dfsctarge header relief heat exch. ctcsa rclfef Scat exch. condensato host exch.
condensate rel lof punp afnfnua flou hest exch. tt>creel relief heat exch. vent FDR aystcn inter-tie CAC systems Loop A drain PIIR-RV 25A RIIR-RV-55A RIIR-V 11A RIIR-RV 36 RIIR FCV 64A RIIR-RV-1A RIIR-V-73A Stilt-V 121 RIIR-V 134A 47 3> 2-6 56 b
6.2-3lp Rtta-V-24A
}O 0
AC Clobo Relief 0 PP.
ltelief 0 YP AC Spring Spring
}O 0
Cate Relief 0
}O 0
Clobe Relief 0 AC pp AC PP AC Spring AC Spriag
}O 0
Clobc Cata 0
AC AC
}O 0
C C
C C
C 2
~
10 F,X3 R}f C
0/C C
AS IS 4
C C
C 35 RH 0
C 0/C AS-IS 3
8 1-I/2 2
Std C
C C
39 Manual C
0/C C
AS-IS 33 Ycs tt Valves R.B.
22 Ycs S
Valves R.B 18 Yes tt Valves R.b.
20 Ycs tl Valves R.b.
22 Ycs tj Valves R.b.
188 Yos N
Valves R.B.
17S Ycs A
Valves R.S.
No 2, 18, 24 No 18 19 tto 18 19
Ão 18 No 18 20 No 18 bo 18 19 No 18 3
}tsnual C
C 0/C AS-IS 2
Std 97 6
No tj Valves R.b No 44 Ycs tt Valves R.S.
No 18 F,X3 Rtf C
C 0/C AS-IS 18 Std 12
'Yes N
Valves R.b.
RIIR Loop 5 poop test i}no discharge header rclfcf hest exch. stean rcltcf puap Adb suctfoa relief host exclt> coa dcnssta puop nfninua flou flush ltno relief hest exch. theresl relief heat exch. vent CAC aystca Loop 5 drain 48 3.2-6 S6 5
6.2-31p RIIR-V 248 RIIR-RV 255 kltR-RV-555 RIIR-RV-S SIIR-V 115 RIIR-FCV-645 jtllR-RV-30 RIIR-RV-lb RIIR-V 735 RIIR-V 1345
'O 0
COte
}O 0
Clubs Relief 0 AC PP Relief 0 PP NO 0
Clobe FO O
Caco AC AC
}O 0
AC Clobe Relief 0 PP Reliof 0 1'P Relief 0 1'P AC Spriag Sprinb SprinS AC AC SprinS Sprfn8 F,X3 R}t F,X3 b}f 38 jof 39 C
C 0/C AS-IS 18 Std 12 Yes tt Valves R.S.
No 2, 1$,
1/>
tto 18 19
Ão 18 19 No 18 19 No 18 30 Yes
}I Valves R.S.
20 Ycs S
Valves R.B.
2 10 C
0 C
C C
C 20 Ycs tt Valves R.B.
C 0
C C
0(C C
AS IS 4
Std 15 Yes lj Valves 5 5, 0
C O/0 AS-IS 3
8 22 Yes jj Valves R,B, j
18 34 Yes tj Vs)vcs R.S.
189 Yes lt Valves K.5.
190 Yes A
Valves R,B.
No 18 19 No 18 19 No 18
>to 18 C
C C
C C
C
}tsnual C
0(C C
AS-IS 2
]a I/2 2
Std
}tsnual C
C 0(C AS-IS 2
Std 44 Yea tj Valves R.5.
TABLE 6i2-l6 (Contfnuad)
DRAFT UNTIL MPPSS REVIEII LINE DESCRIPTION CAC Dfvfsfoa 1 dfachorge co dryvell 96
- 3. 2-17 56 B
GAG-V-2 6.2-31g HO 0
DC Care EIIO 0
AC Globe 31 Hanuel C
C 0/C C
2-Sc4 1/2 31 Hanual C
0 0/C AS 1$
4 Std 4
ies A
Valves R.B, No 17 CAC Dfvfsfoa 2 succfon froa dryvell 91
- 3. 2-11 56 B
CAG-V-15 6.2-31$
CAC-FCV IB HO 0
DC Ceca EIIO 0
C 0/C C
2-St4 1/2 37 Hanuel c
c 0/c hs-Is 4
scd 2
Tes A,S Valves R.B, No 11 CAC Dfvfsfoa 2 dfscbarge co dryvel1 9$
3r2-11. 56 B
CAG-V-ll 6r2-3lg CAG-PCV-2B HO 0
C 0/C C
2-Scd 1/2 10 37 Hanual C
C 0/0 AS-IS 4
Std 8
Yes A
Valves R.B.
No 11 CAC Dfvfsfoa 1 suetfon Pron dryvell 99 3e 2-17 56 B
CAC-V-6
- 6. 2-318 IIO 0
DC Caco EIIO 0
C 0/C C
2-Scd 1/2 31 Hanual C
C 0/C AS IS 4
Scd 4
Tea Ayg Valves RoB ~
No 11 1
CAC Dfvfsfoa 1 dfsebnrge to vetvoll CAC Dfvfafoh g dfscIIscgc )0 vstvcll 10) 3,2-17 50 Sol 31S CAG;FCV 44 a
CAC-V-13 Chr FCF-4S 102 3,2-17 56 B
GAG V-i 6.2-3lg HO 0
DC Caco EIIO 0
AC Clobo
}ID 0
QQ Caco 73I0 0
C 0/C AS-IS 4
Std 3
Ves A
Valves R.B.
No 11 37 Hnnuel C
C 0/C C
2-Scd 1/2 37 7buIual 0
C 0/0 c
2-Scd 1/2 37
.'Iahual 0
0 0/0 AS 1S 4
Scd 7
Too h
Volvoo Rigi No l1 CAC Pfv)sfoa l euccfon trca vccvoll 104 3)2-17 56 g
GAC V"17 Sel 31S CAG FCV~
3S HO 0
DC Gato 73IG 0
hC fIlobo 37 Hanucl C
37 Hanuel C
C 0/0 C
0/C 2-Scd l/2 AS-IS 4
Scd 5
Voo h,g Valves Regi No 17 CAC Dfvfaloh 1 auction froa vecvell 10$
3.2-17 56 7I CADY-S 6.2-31g HO 0 M Cata EIIO 0
hC Globo 37 Henuel C
C 0/C C
2-Std 1/2 37 Hanuai 0
0 0/0 AS 18 4
Scd l Yoo A,S Vaiveo S S.
ga 17
4
. ~
a
..~ ~,e-at~ ~
r
~
~ u>>
~ ~ ~
~
~
~ i.
~
~
YABLE 6.2-16 (Conttnued)
~,
DRAPE HHTIL HPPSS REVIEW LINE OESCRIFTIOH RHR Loop A Sup-35 pression Pool Suet fun RIIR Loop B Sup-32 pression Pool huctfon RIIR Loop C Sup-36 pression Pool Suction
- 3. 2-6 56 B
RHR-V-4A N 0
AC 6.2-31n Cate 326 56 B
RIIRV4B N 0
RIIR-V-4C MO 0
AC 6.2-3ln Cats AC 46 Manual 0
0/C 0
AS-1S 24 Std 2
Yes N
Valves R.B.
Hu 18 AC 46 Manual 0
0/C 0
AS-1S 24 Std 2
Yea N
Valves R.b.
No 18 AC 46 Manual 0
0 0
AS-1S 24 Std 2
Ycs H
Valves R.b.
No 1S RIIR Loop At 111 heat exch. stean rclIc!
condeneatc pot drain condensate pot drain IOlb Loop BI 118 heat exch, stean rcllof condensate pot drain condensate pot 4talb 3.2-6 56 B
6.2-314 3.2-6 56
- 6. 2-3ld RIIR-RV-95A RIIR-V-124A RHR V 1241 RIIR-RV 95B RIIR-V~
123h RIIR-V" 058 MO
~ 0 Cats MO 0
Cate AC AC RclSet 0 PP N
0 Cato MD 0
Cote AC Rolls! 0 PP Sprfnl C
C
~
C AC 39 Manual C
C C
AS-18 Sprfnl C
C 0
AC 39 Manual C
C C
Manual C
0 C
C C
AS-1S 10 1-Std 1/2 1-Std 1/2 10 1-St4 1/2 1-St4 1/2 24 Yes S
Valves R.b, 11
'Yes H
Valves R.b, No 1S 19 No 18 21 Yca S
Valves R.B.
Yeo H
Valves R.b.
No 18 19 No 18 14 Ycs H
Volves Il,l.
Ilo $ 8 12 Yes H
Valves R.b.
No 18 RIIR Inop Cl 26 punp teat llno discharge header rcllvC pusp C suction rolicf pusp nlnlnun Clou 3>>2-6 55 R
6e2 31E R)IR-V-21 RIIR-RV 25C RIIR-RV SBC RHR-V-640 Suppression Pool 25h 3.2-6 SS l RIIR-V Spray Loop A 6.2-3lh 2)A IIO q
Clubs RolSoS 0 Relief 0 PP N
0 AC Clobo MO 0
C C
AS-'lS 18 Std 34 Yes lf Valves R.b.
No 1S 30 Yep M
Valves R.b, 37 Vco M
Valves R.b>
G 9
C Ito 18 19 Mo
$ 8'9 No 18 Sprig C
C C
~ ~
hc 7,23 Iol C
C 0/C AS-1S 6
10 5
Ycs lf Valves R.bo No 2, 18, 24 AC 38 RM 0
C 0/C AS-IS 3
8 30 Ycs M Valves R.bo Sup prcaefon Pool 25b 3.2-6 56 B
RIIR-V-Spray Loop b So2-3lh 278 MO 0
C 0/C AS-1$
4 10 6
Ycs I Valves R.b.
No 2 ~
18, 24
'7 Z
~r> ~ &
f4 4 H >>r 4w
~
~
TABLE 6.2-16 (Continued)
I'J H>
I ~
Dying SHYg, QPPSS REV)Etf lu LlNL DESCRIPTlON RB co Nccvell Vacuun Breakers 119 3> 1-15 56 B
CSP-V-9 6.2-31%
CSP-V-10 36 B
CSP-V-5 CSP-V-7 CSP-V-4 CSP-V-3 RS co Qecvell 66 3.2-15 Vecuun Breakers 4
6.2-31b Nctvell Vcntfl-6.2-31%
ation Supply AO 0
Sutfy PC 0
Cbcck AO 0
BucCy PC 0
Obcck AO 0
Butfy AO 0
Butfy Spring hfr Process Process Spring Afr 40 RM Air Aft'pring F,B,X RM Spring F,B,X RM Process Process -
RM C
C C
C C
C 0
C C
0 24 24 4
24 4
Yes A
Valves R.b.
No 26 7
Yes h
Valves R.b.
No 17 10 Yes A
Valves R.B.
No 16 C
C C
0 C
C C
C 24 4
14 No A
Valves R.b No 24 4
17 No h
Valves R.b.
No C
C C
0
~
24 4
1 Yca h
Valves R.B.
No 17 RB co Nccvell 67 3.2-15 Vacuun Breakers 4
6>2-3lf Hccvell Vcncff-6>2-31%
ation Exhaust 56 b
CSP-V-6 CSP-V-B CEP V-4A CEP-V-3A CEP V-4b CEP-V-3B AO 0
Butfy PC 0
Clicck AO 0
Butfy AO 0
Butfy AO 0
Ceca AO 0
Cato Air Spring F,B,Z RM Spring hir 40 RM Process Process RM Spriug F>B,Z RM Spri>>g F,B,Z RM Alr Spring F,B,X RM C
C C
0 C
C C
0 C
C C
C C
C C
C C
C C
C C
0 C
24 4
24 24 4
B Yes A
Valves R.B.
No 17 16 Yes A
Valves R b No 16 10 No h
Valves R>bi No 1
10 No h
Valves R.b.
No, 2
1 12 No h
Valves R.b.
No 24 4
12 No h
Valves R.b.
Ho Dryvell Ventfl-aclon Supply Dryvell Ventil-ationn M41ust 53 3.2-15 56 B
CSP-V-2 6.2-3lb CSP-V-l 3
3r 2 15 55 b
CEP"V-lh 6.2-31$
CEP-V-1A CEP-V"1B CEP V 2S AO
.0 Butfy AO 0
Butfy AO 0
butfy AO 0
Butfy AO 0
CAco AO 0
Cata hlr Atr Air Afr Spring F,B,Z RM Sprlug F,B,Z RM Sprfng F,B,X RM Spring
~ F,b,g R.'f Spring F,S,Z RM Spring F,B,Z RM C
C C
C 0
C C
0 C
C C
C C
C C
C C
C C
C C
C C
C 30 4
1 No A
Valves 30 4
4 No A
Valves 30 4
12 No h
Valves 30 4
S No h
Valves R.br R,b, r R>b>rR.l'o 17 No 17 2
1 12 No A Valves R.B.
No 2
1 8
No h
Valves R.b, No
~ e TARLE 6.2-16 (Continued)
DRAFT UNTIL MPPS5 REVIE})
LINF. DESCRIPTION RCC Inlet Deader S
3o 2-14
$ 6 5
RCG-V-104 6+2 3lt RCG-V-5
})O Caco HO Cato 0
AC c/g)d 0
0 AC )I~ -
0 0
AS-IS IO Std S
No M
Valves R.b.
No 17 AS-IS 10 Scd 3
$ 6 5
RCG-V-21 6i2-31o RCG-V-40 NO Cata NO Cato 0
AC I
AC AC i/5gf 0
0 AC r/+5}
0 0
AS-IS 10 Std 3
No ll Valves R.b.
No AS-IS 10 S cd Suppreaaloa Pool Cleanup Suec}on 100 3+2 12 6.2-311 S6 5
IPC-V-IS3 FPC-V"}$4
}I Cato.
)I)
Cato 0
AC 0
AC AC F,b RH C
C C
AS-IS 6
Std 7
AC Feb RH C
C C
AS IS 6
St4 2
No C}
Valves Robe bo 11 Suppress)on Fool Cleanup Return 101 3,2-12 S6 5
FPC-V-1$ 6 6i2 3lo FPG-V-149
)Q Caco Clobo 0
AC 0
AC F,b RH C
C C
AS-IS 6
Std 3
No Valves 5'C No ll Wl Pron Reactor 14 3.2-11 A RUCU-V-I NO 6+2-315 Cata RNCU-Y"4 )lO Cato I AC AC hyJy RH 0 0 +)i~e Cgt } e DC A,J, RH 0 0 42- ~ f:,./,}4 C AS-IS 6 Std No M Valves Rb No C AS IS 6 St4 4 RRc Punp A seal Mater l 4)h 3 I2~3 36 5 RRC Ve .6,2-$la IIA RRC V~ }6A A~ac}) }lO Cate I 1'rooooo 0 AC Pcoooos << 0 0 AC Af 0 0 C 3/4 Std o llo }) Valves R 5, No c As-Is 3/4 sc4 RRC Fusp 5 coal uacer 435 3e2 3 56 b RRG.V 6e2-315 135 RRG-V 165 })0 Cate 0 AC Cbeck I Process Process 0 0 ~/r AC PT 0 0 C ~ 3/4 Std No N Valves R.b, No C AS IS 3/4 Std 2 RRC Sanplo Lfno 694 3 2 3 5$ h RRG-Y"19 6,2-314 RRG-V-20 SO'lobo AO Globe I AC RH C 0 C C Spring B,C, Dep Sprinb 5,C, RH D,P ' C 3/4 eS C C 3/4 St4 No }} Valves T.bi .0$
TABLE 6o2-16 (Contiauc4) c DRAFI UNTIL NPPSS RZVIEN LDDE DESCRIPTION Dryuell Equipaeat 23 3.2 9 56 b FDR-V-19 AO 0 Drain
- 6. 2-31k Gate EDR-V-20 AO 0
Cats hir Air Spring t,b RH Spring t,B RH 0 0 C C 0 0 C C c 3 Std 2 No M Valves R.b; No 17 3 Scd 4 Dryucll Floor Drain Decontaainacion Solcn. Supply Header Decontanination Solcn. Rccurn Header 24 3.2-10 56 B FDR-V 3 AO 0 6.2-31k Cate PDR-V"4 AO 0 Cate 94 3.2-10 NA 95
- 3. 2-10 NA b
hir Spring t,b RH Spring t,b RH 0 0 C C 0 0 C C R.bo No 17 4 blanked R,bo Close Blanked Close R.be No Std 2 No ll Valves 3 Scd 3 Clh for Safety Relief Valve hcc<<cuiators S6 3 2 21 56 b CIA Y 21 Chock 0 Process Process C C C 6.2-3lc CIA-V-20 NO 0 AC AC 41 Hanual 0 0 0 Clobo 3/4-AS-IS 3/4 Std 3 No A Valves R.b. No 17 10 CIA Line.A for 89B 3.2-21 56 b Clh-V-31A Check 0 I'rocess Process ADS Accuaulators 6.2-31c CIA-V-30A NO 0 AC AC Globo 42 Heauel C C C 1/2 0 0 0 AS-Ig I/2 Std S No A Valves R.b,- No 17 15 CIA Line B for ADS hccu-ulacors 91 3.2 21 S6 b CIA-V-31b Check 0 Process Process ~ C C C 6.2-3lc Clh-V-30B lm 0 AC hC 42 Hanual 0 0 0 Globe I/2- 'S-IS 1/2 Scd 2 No A Valves R.b, No 17 CRD Iaserc Lines 9 3.2 4 56 b (185 separate lines) CRD Nithdraval 10 3.2-4 SS B lines (185 separate liaes) Sco Note 4 Seo Note 4
Thgl.E 6.2-16 (Cont fnued) DRAFT UNTIL MPPSS REVIEM LlNE DESCklPTION hir llno for 42d 6.2-3lr 56 B PI-VX-tasting kRR-V-50h 3.2-6 42d PI-VX-216 Globo 0 ".Hanual Hanual Globo 0 Manual Hanual LC LC LC ,LC LC LC <7 Ho h Valves R.B. No 25 <7 hir line for 54BE 6.2-3lr 56 B PI-YX-casting RHR-V-41B 3.2-6 54BE PZ-VX-218 Globe 0 Hanual Hanual Globe 0 Hanual Hanual hlr linc for 62E 6.2-31r 56 B PI-VX-62E Clobo 0 Hanual )hnual testing RHR-V-41C 3.2-6 PI-VX-220 Globe 0 Hanual Hanual Alr line for test,lng LPCS-V-6 Air linc for testing NPCS-V-5 78d 6.2-31r 56 B PZ<<VX-78d Clobe 0 Hanual Hanual 3.2-7 PI-VX-222 Globo 0 Hanual Hcnual 78c 6.2-3lr 56 B PI-VX-78o Globe 0 Hcnual Hanual 302-7 PZ-VX-223 Clobo 0 Hcnual Hanual hlr llno for 54ha 6.2-3lr 56 B PI-VX" casting RC)C-V-66 3.2"8 54Aa PI"VX-217 Globo 0 Hanual Hcnual Globe 0 Hanual Hanual hfr linc Eor 69c 6.2-3lr 56 B PI-VX-69c Globe 0 Hanual Hanual teatfng RRR-V-50B 3.2-6 PI-VX-221 G)obe 0 Hanual Hanual hlr linc for 61E 6.2-3lr 56 B PI-VX-61E Globo 0 Hanual Hanual testing RNR-V-4lh 3.2-6 PZ-VX-219 Globo 0 Hanual Hanual LC LC LC LC LC LCi LC LC LC-LC LC LC 1C LC LC LC LC LC LC LC LC LC 1C LC LC LC LC LC LC LC LC LC LC LC LC LC LC LC LC 1C LC LC <7 No h Valves R.B. <7 <7 No h Valves R.B. <7 Ho 25 Ho 25 <7 Ho h Valves "R.B. No 25 <7 <7 Ho h Valves R.B. <7 No 25 <7 No h valves <7 R.B. Ho 25 <7 No h Valves R.B. Ho 25 <7 <7 Ho h Valves R.B. No 25 <7 hir linc for casting MM-DM vccuias relic E valves 82a 6.2-3lr 56 B CAS-Y-453 9.3-1 CAS-CVX-82e SO 0 Clobe Cbeck 0 AC Spring 44 Process Process C C C C C C C 1 <5 5 No h Valves R.B. No P~ hlr linc Eor maintenance 93 9r3-I 56 B 6.2-3lt Sh-V"109 Pipe I Cap Cate 0 Hanual Hanual 0 C C LC LC LC Ho h Cap Valve S,B, Ho Tlp lines 27a-e 54 - C5lJ004 C513004 SO 0 AC AC g P Ball Shear '0 - Explosive 43 IOf C C C C 3/8 <5 2 No h Valves R.B. Ho 2. g 0 0 ~ 0 0 3/8
0 I'
~~) ~ 1 0 ~ ~ 'I8 ~ ~ ~ lf 0 ri ~ s <<1w ~ a qy) ~ sgtppygp ~ ~ ~ ~ 2 TABLE 6.2-16 (Contfnucd) Llgg DESCRIPTION %4PT fgfPIL lyPSS REVIKN ~ g Radfatfon Honitor 85b 6.2-3ls S6 {S-SR-20) Supply 3.2-15 1 foe SO 0 Cata Pf-% ~ISO 0 Cato AC Sprfng AC Spring 00001c5 0 0 C C 1 cS No h Valves Rambo No 25 Radfatfon Honfcor 73a 6.2-3le 56 (S-SR-2l) Supply 3.2-1S lfne Radfatfon Nonfcor 29/ 6.2-31s S6 (S-SR-21) Return Q, 3.2-15 1fno B Pr-V/f-ZSGSO 0 Cata FI Q, FS7SO 0 Cata FI fh 2~60 0 PMt All Inacriaaenc lfnes Iron reactor 55 h KF 0 Chcc'k Globo 0 Redfatfon Honftor 72I 6.2-3ls 56 B ~ (y'660 0 (S-SR-20) Return 3.2-1S + 7 ~ Caco lfne 'hcck 0 AC Process Process AC Sprfng AC Sprfng AC i( Sprfng 47 Procoas Process Sprfng KF Hanuel fcmual i/g Sprfng Af 0 0 C C 0 0 C 0 0 C 0 0 . C 0 0 C C 0 0 C 0 0 0 0 0 0 1 <5 1 1 cS 1 <5 1 <5 1 3/4-6 1 3/4 << 6 1 No A Valves R.B. No 25 No A Valves R.B. No 2S 'No h Valves R.B ~ No 2$ Valves R.BE No 27 All Inatrunent lfnes Iron prf-cary contafnnent 56 B KF 0 Sprfng Check Globo 0 Hanual Hanual 0 0 0 0 0 0 1 4 I-1/2 1 6 1-I/2 Volvea R.B. No 27 Inacruaenc lfncs (Hydrogen nonftors) rccurn to contafn-nenc 3,2 15 56 B Check 0 Process Process Globo 0 Hanual Hanual C C 0 0 0 0 Ves A, Valves-R.B. '1 No 27
t
tZ g>C TABLE 6.2-16 PRIMARY 'CONTAXNMENT..AND.REACTOR VESSEL XSOLATXON 'SIGNAL'ODES FOR TABLE 6.2-16** S icinal Descri tion Reactor vessel low water-level (Trip 3) (A scram occurs at this level also. This is the higher of the three low water level signals). D* Reactor vessel low water level (Trip 2) High radiation - Main steam Line break Main steamline (steamline high space temperature or high steam flow). High drywell pressure (core standby cooling systems are started) K* Line break in cleanup system high space temperature. Line break in RCXC system line to turbine (high RCXC pipe space temperature,'igh steam flow, or low steam line pressure).. Line break in RHR shutdown piping (hi suction flow) Low main steamline pressure at inlet turbine (RUN mode only). These are the z.solation functions of the primary contain-ment and reactor vessel i'solation system; other functions are given for information only.
- See notes 30 through 46 for isolation signals generated by
, the individual system process control signals or for remote-manual closure based on information available to the operators. These notes are referenced in the "isolation signal" column.
TABLE 6.2-16 (Continued) Descri tion S icina1 S Low drywell pressure U*" High reactor vessel pressure High temperature't outlet of cleanup system non-regenerative heat exchanger Y Standby liquid control system actuated Reactor building ventilation exhaust plenum high radiation Remote manual switch located in main control roog G* Low condenser vacuum H* Turbine Building high temperature T* High leakage flow X* "K" plus RHR/RClC equipment area high temperature Xl~ High drywell pressure and low reactor pressure X2" RHR equipment area high temperature X3* Reactor vessel low water level (Trip 1) E.* R. A ~('A'le Q<~ a-Up S9 S~~ g~@~ QiF's.-gg,g ,<~<f~~~ > These are the isolation functions of the primary contain-ment and reactor vessel isolation system; other functions are given for information only.
0 I
I, ~ TABLE 6.2-16 (Continued) ABBREVIATIONS/LEGEND
- Valve, MO PC EHO SO Type air operated motor operated positive closing electro-hydraulic operated solenoid operated Location I
inside containment 0 outside containment Power to Open/Close DC I~ ~Process, pro PP spr AC electrical power DC electrical Power process flow process fluid overpressurization spring Normal Position 0 open C close Process fluid W . water A air S steam H hydraulic fluid Termination Zone TB RB Rad W SB turbine building reactor building radwaste building service building
0 )
3'2/P-2 R TABLE 6.2-16 (Continued) NOTES FOR TABLE I C testing is discussed in Figure 6.2-31 which shows the ation valve arrangement. Unless otherwise noted (see s 4, 27, 28,29) all valves listed in Table 6.2-16 are C tested. Type isol note type l. 2. 3. 4 Main steam isolation valves require that both solenoid pilots be de-energized to close valves. Accumulator air pressure plus spring set act together to close valves when both pilots are de-energized. Voltage failure at only one pilot does not.cause valve closure. The valves are designed to fully close in less than 10 seconds. Suppression cooling valves have interlocks that.allow them to be manually reopened after auto'matic closure. This setup permits suppression pool spray, for high drywell pressure conditions, and/or suppression water cooling. When automatic signals are not present, these valves may be opened for test or operating convenience. 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. The isolation provisions for the CRD lines are commen-surate with the essential requirement. that the control rods are inserted on a scram. Isolation of the hydraulic lines is provided by check valves 115 and 138 'nd solenoid valves
- 120, 122, and 123 on the hydraulic control units (HCU) and by air operated valves F010, F011 on the scram discharge header (see Figures 4.6-5a and b).
The HCU's and scram discharge headers as well as the. hydraulic lines themselves are Seismic I., and are qualified to the appropriate accident environment. The failure and scram p'osition of all power operated valves are compatible with system isolation and, at the same time, rod insertion on a scram. Addition of power ~ operated isolation valves on the hydraulic lines them-selves could prevent control rod insertion. Manual isolation valves 101 and 102 allow for further isolation if it becomes necessary. The hydraulic lines are'mall and terminate in the reactor building which is served by the standby gas treatment system. The hydraulic lines and their manual isolation valves in the scram discharge header and its air operated valves are code group B.
11 ~ i ~ I4 < v WNP-2 tq g~S TABLE 6.2-16 (Continued) The hydraulic control uni4..(HCU)..is.a General. Electric factory-assembled engineer'ea*"'mo'dule of va10'es',
- tubing, piping, and stored water which controls a single control rod drive by the application of precisely timed sequences of pressures and flows to accomplish slow insertion or withdrawal of the control 'rods for power control,
- and, rapid insertion for reactor scram.
Although the hydraulic control unit, as a unit, is field 'installed and connected to process piping, many of its internal parts differ markedly from process piping com-ponents because of the more complex functions they must, provideo Thus, although the codes and standards invoked by Groups A, B, C, and D pressure integrity quality levels clearly apply at all levels to the interfaces between the HCU and, the connecting conventional piping components (e.g., pipe nipples, fittings, simple hand valves, etc.), it is consid-ered that they do not apply to the specialty parts (e.g., solenoid valves, pneumatic components, and instruments). The design and construction specifications for the HCU do invoke such codes and standards as can be reasonably ap-plied to individual parts in developing 'required quality levels, but these codes and standards are supplemented with additional requirements for these parts and for'he remaining parts and details. For example,
- 1) all welds are penetrant tested (PT),
- 2) all socket welds are in-spected for gaps between pipe and socket. bottom,
- 3) all welding is performed by qualified welders, and
- 4) all work is done per written procedures.
Quality Group D is generally applicable because the codes and standards invoked by that group contain clauses which permit the use of manufacturer's standards and proven design tech-niques which are not explicitly defined within the codes of Quality Group A, B, or C. This is supplemented by the QC techniques. The CRD lines will be included in the type A test leakage since the reactor pressure vessel and the nonseismic por-tions of the CRD system are venked during the, performance of the type A test. The CRD insert and withdraw lines are compatible with the criteria intended by 10CFRSO, Appendix J, for type C testing, since the acceptance criterion for type C testing allows demonstration of fluid leakage rates by associated bases.
TABLE 6.2-16 (Continued) Alternating current motor-operated valves required for. iso~-."'4:,oa.:functio'ns are;powered from the AC standby ""'"'ower buses. Direct current operated isolation valves are powered from station batteries. All motor-operated isolation valves remain in the last position upon failure of valve power. All air-operated valves close on motive air failure or in the safest position. The standard minimum closing rate is 12 inches of nominal valve diameter per minute for gate valves'nd 4 inches of valve stem travel per minute for. globe valves. -For. example, a 12 inch gate valve will close in one minute. Reactor building ventilation exhaust plenum high radia-tion signal (2) is generated by two txip units; this requires one unit at high trip or both units at down scale (instrument failure) trip, in order to initiate isolation. Primary Containment and Reactor Vessel Isolation Signals (PCRVIS) are indicated by Letters. Isolation signals generated by the individual system process control sig-nals or for remote manual closure based. on information available to the operator are discussed in the referenced notes in the "isolation signal" column. Normal status position of valve (open or closed) is the position during normal power operation of the reactor (see Normal Position column). The specified closure rates are as required'or con-tainment isolation only. Reported times are in. seconds. All isolation. valves are Seismic I. Used to evaluate primary containment leakage which may bypass the secondary containment emergency filtration system. See 6.2. 3. 2. Reported sizes are the valve nominal diameters in inches. Size indicated is containment side of relief valve when relief valve size is not equal on both sides.
0 .I
HNP-2 TABLE 6.2-16 (Continued) 15 ~ 16. 17. 18. , Leakage control system provided, see,6...7,~,,+., Bypass leakage of secondary containment is not con-sidered during design basis LOCA, see 6.2.3.2. Valve operability will be improved because the environ-mental conditions are better outside the primary con-tainment from the standpoint of humidity, radiation, pressure and temperature transients, and post-LOCA pipe whip and jet impingement. These lines connect to systems outside of the contain-ment which meet the requirements for a closed system as set by NRC SRP 6.2. 4, Section IX, paragraph 3e. These systems are considered an extension of the primary containment. Any leakage out, of these systems-will be processed by the standby gas treatment. system. 19. Relief valve setpoint greater than 77.5 psig (1.5 times containment design pressure). 20. Relief valve 'setpoint is 75 psig. 21. 22. Cannot ae reshut after opening without disassembly. See 6.2.4.3.2.2.1.2 23. See 6.2.4.3.2.2.2 24. Due to redundancy within the emergency core cooling
- systems, some subsystems may be secured during the long, term cooling period.
Xn addition RHR loops A and B have several discharge paths (LPCX, Drywell Spray, Suppression Chamber Spray, Suppression Pool Cooling) which the operator may select during the 30 day post-LOCA period.. 25 '6. Applicable portion of the flow diagrams 3.2-6, -7, -8, ~ -15> to be updated to 'reflect the configurations shown on Pigures 6.2'31r and -31s. Qrr4l Q.5. I The disc on the check valve is maintained in the close position during normal operation by means of a spring actuated lever arm and magnets embedded in the periphery of the disc. The magnetic and spring forces maintain the disc shut until the differential force to open the valve exceeds 0.2 psid. The check valves have position
0 ,I
~ ~ TTNP-2 ~ ~ ~ <> ~wy'sf'ABLE 6.2-16 (Continued) ~~yam indication liighus whi'ch can alert the operators to the fact that the check valve is not fully closed. The operator can then remotely shut the valve by means of a pneumatic operator. The operating switch return to neutrral so the vacuum breaker function will Quality Class I. not be impaired. The air supply to th s o ese valves is 27. Instrument 1'nes that penetrate primary'ontainment: conform to Regulatory Guide 1.11. The lines that connect to the reactor pressure boundary include a restricting orifice inside containment S tegory I and terminate in instruments that. are Seismic Category I. 'he instrument lines also include manual isolation valves and excess fl h ow c ack valves equivalent (see hydrogen monitor return lines). the inta r'hese panetrations will not be type C t d g ity of the lines are continuously d e asta since stratad durin lant o y amon-g p an operations where subject to reactor operating pressure. In addition all 1'znes are subject ype est pressure on a regular interval. Lfeaktight integrity is also verified with complet f won o well as b 1 and calibration surveillance a t y visual inspection during daily o arator ~ ~ ac xvxtxes as patrols as applicable. x, y operator 28. Penetrations Z-76 and X-77 contain li h draulic c ain xnes for the y ic contro1 of the reactor recirculat'l son ow ese lanes contain corrosive hydraulic control valve. used used to position the reactor recirculation flow These lines inside of the containment: are Seismic Category I and Qualit Gr u ax e closed automatic isolation valves outside the o h'd 'containment which receive n ig rywell pressure. an automatic isolation signal. Crita These lines meet. the recpxirement f G o eneral Design erron 57 and therefore reauire onl sin le isolation valves outside of the T ay are designed to Seismic Category I, Code Group B and the following criteria:
WNP-2 x3 g>K TABLE 6.2-16 (Continued} "--~-.~:. <<-"<<do-not communicate:width'"either the reactor coolant system or the containment atmosphere, b. are protected against missiles and pipe whip, c. are, designed to withstand temperatures at least equal to the containment design temperature, d. are designed to withstand the external pressure from the containment structural acceptance
- test, and e.
are designed to withstand the loss-of-coolant 'accident transient and environment. Even if the failed closed valve were to not shut there will be no leakage of containment atmosphere through the hydraulic control lines since the piping inside the primary containment remains intact. There are no active component failures which would compromise the integrity of.the closed system inside the primary con-tainment. Integrity of the closed system inside the primary containment is, essentially, constantly moni-tored since the system is under a constant operating pressure of 1800 psig. Any leakage through this system would be noticed because operation would be erratic and because of indications provided on the hydraulic control unit. In addition, in order to perform type C tests on these li'nes, the system would have to be dis-abled and drained of the corrosive hydraulic fluid. This is considered to be detrimental to the proper operation of the system in that possible damage could occur in establishing the test condition or restoring the system to normal. These lines and associated isolation valves should there-fore-be considered to be exempt from type C testing. 29. 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 closely described by General Design Criterion 54, which states in effect that isolation capability of a system should be commensurate with the safety importance of that isolation. Furthermore, even though the failure of the TIP system lines presents no safety consideration, the TIP system has redundant isolation capabilities.
A r
~ WNP-2 TABLE 6.2-16 (Continued) The safety features have been reviewed by the NRC for BWR/4 (Duane:Arnold), BWR/5 (Nine Mile Point) and BWR/6 (GESSAR), and it was concluded that the design of the containment isolation system meets the objectives and intent of the General Design Criteria. Isolation is accomplished by a seismically qualified solenoid-operated ball valve, which is normally closed. To ensure isolation capability, an explosive shear valve is installed in each line. Upon receipt of a signal (manually initiated by the operator), this explosive valve will shear the TXP cable and seal the guide tube. When the TXP system cable is inserted, the ball valve of the selected tube opens automatically so that the probe ~ and cable may advance. A maximum of five valves may be opened at any one time to conduct calibration, and any one guide tube is used, at most, a few hours per year. Xf 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 with-drawal. Xf 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 WNP-'2 TXP system design specifications require that the maximum leakage rate of the ball and shear valves shall be in accordance with the Manufacturer's Standard-ization Society (Hydrostatic Testing of Valves). The ball valves are 100% leak tested to the following criteria by the manufacturer: Pressure 0 62 psig Temperature 340 P Leak Rate 10 cm /sec -3 3 A statistically chosen sample of the shear valves is tested by the manufacturer to the follows.ng criteria: Pressure 0 - 125, psig Temperature 340 F Leak Rate 10 cm /sec STP -3 3
~5 gran TABLE 6.2-16 (Continued) The shear valves have explosive squibs and require ~.".- ""-"""'esting to destruction. They cannot therefore be 100% tested. As stated
- above, the penetration is automatically closed following use.
During normal operation the penetration will be open approximately eight hours per month to obtain TXP information. Xf' failure occurred such as not being able to withdraw the TXP cable, the shear valve could be closed to isolate the penetrations. Installation requirements are that the guide tube/penetration flange/ ball and shear valve composite. assemble not leak at a rate greater than 10 4 std cc/sec at 80 psig. Further leak testing of the shear valves is not recommended since destructive testing would be required. Leak testing of the ball valves also is not recommended since the guide tube terminates in a sealed indexer housing which is kept under a positive pressure .by a nitrogen or.air purge. The purge make-up will be indica-tive of the system leakage. Note that the TXP ball valve is normally closed and thus is a part of the leak-age barrier being monitored. Consequently, the personnel exposure required to conduct type C tests from inside the containment is not warranted. 30. System is initiated after a LOCA. Isolation valves will auto close on the following high leakage conditions: a. 5 PSI between main steam isolation valves, 60 seconds after system initiation b. High flow from main steam line to low pressure
- manifold, 150 seconds after system initiation 31.
c. Inboard main steam isolation valve opened, 'after system initiation PCRVXS is not desirable since the feedwater
- system, although not an ESF system, could be a significant source of make-up after 'a LOCA which is not concurrent with a seismic event.
Feedwater check valves on either side of the containment provide immediate leak isolation, if required. The feed-water block valves can,
- however, be remote-manually closed if there is no indication of feedwater flow (see 6.2.4.3.2.1.1.1).
WNP-2 'ABLZ 6.2-16 (Continued) 'bc, c 8ICtl'iYI 32. he RCIC system at most onl during the first seve=al hours after a LOCA. T e RCIC system wit maihC be xnxtxated by ow water level (B signal) and subsequently will be automatically tripped by one of the turbine shutdown si nals listed b low. The operator upon receiving indication t at the RCIC system is no longer operable should complete isolation of the system by remote-manually shutting the isolation valves which have not been automatically shut. Also t4 oP<<Nor will iVi<<<t"~ Relic. eq~<e~ on a hgih level alarm in ting appropriate. Reactor Bu'ildln9 50~P'utomatic shutdown of the RCIC turbine occurs upon receipt of any one of the following signals: a. turbine overspeed b. high water level in the RPV c. low RCIC pump suction pressure d. high turbine exhaust pressure e. closure of steam supply valves Automatic closure of steam supply valves to the turbine occurs upon receipt of any one of the following signals: a. high flow in steam supplg line b. high area temperature c. low reactor pressure of 50 psig d. high pressure between tu bine exhaust diaphragms The low reactor pressure signal of 50 psig is expected to occur almost immediately after a design basis LOCA and within several hours for a small LOCA. Leakage from this system, e.g., packing gland, pump seals,'s expected to be negligible because of the small leakage rates expected and because of the short operating time (see reply to Question 212.003 for an estimate of maximum leakage rates expected and the radiological consequences)
NNP-2 TABLE 6.2-16 (Continued) 33. - - turbine 4lirottIe valve, closes loilo~ 34. whrHle mal<<. closes 4oiioi"'"J 35. The RCXC minimum flow valve is open only between the time of system initiation and the time at which the system flow to the RPV exceeds 40 GPM. The valve is shut at all othe'imes. RCIC-U-19 auto closes ~L.v/4<<~~c.. turbine trip (see 32~~)'. 5houQ a leak occur when the valve. is oP++z <4 wi'll be, detected by a high level olarm in ftie appropriate~ pzaefor Build<+ The RCXC injection valve is open only during RCXC turbine operation. Injection line, check valves on either side of the containment provide immediate leak isolation, if required. RCXC-V-13 auto closes ~ when 6C "><"ii< a turbine trip (seev 32'W~). The RCXC steam exhaust valve, RCIC-V-68, is normally .open at all times. Should a leak occur, it would be ~h h' tron system. ~ 1 (g ~Mote 3~ 36. The RCIC vacuum pump discharge
- valve, RCXC-V-69, is
. normally open at all times. The valve could be remote-manually closed by the operator upon indication that .vacuum (annuniated in Main Control Room) can no longer be maintained in the barometric condenser. 37. 38. System isolation valves are normally closed. Syst: em 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 another, it would be detected in the main control room and the process loop in service could be shutdown. The minimum flow valve for an FCCS pump is open only between time of ECCS initiation and the time at. which the system flow to the RPV exceeds 640 gpm. The valve is shut at all other times. gglve i s open on<< >turing shddowvi gatv'e. tMsi~ion conqrol r oom iI o roofide. oPerakor eon<triviaHioo o valve s a l 4 kuS. 0gchgr g4A 4hth Vabh'-hS hhPehhh fi VIII~ 5~ho>41 +L a~<<i .,+ ~gg appropriGtg %medor ~~'le'~g ~u P'
0 A
WNP-2 TABLE 6.2-16 (Continued) pI'I pkly, jp %(l( 40. Normally closed. Signalled to open if reactor building pressure exceeds wetwell pressure by 0.5 psid. Valves automatically reshut when the above condition no longer exists. Operakov-fo Use. yale. Posif'ion indicator ag confirmation 04 valve. st0.~4s. 41.. Indication of containment air compressor discharge header pressure and a low pressure alarm exist, in the main control room. The operator can xemote-manually shut valve CIA-V-20 should the containment air compres-sors become unavailable. The isolation check.valve, CIA-V-21, provides immediate isolation. 42. Indication of nitrogen bottle header pressure and a low pressure alarm exist in the main control room. The operator can remote-manually shut valve CIA-V-30(A,B) should the nitrogen bottle bank pressure decrease below the alarm setpoint. The isolation check valves, CIA-V-31(A,B) provide immediate isolation. 43. The operator's indication that remote-manual closure of the TIP shear valves is required, is failure of the TIP ball valves to close as moni'tored on Panel S. 44. Normally closed. Opened only when testing wetwell to drywell vacuum breakers. 45. o lly los V ve i 'nte ocke ith C sy tern, o p n wh n C s s 's 1 e i o r x a o h t w C y em i sh t Th C s m xs s d re an ens o he ima cont Und r de i, n vleuJ. 46.. These valves are the ECCS suction and discharge iso-lation valves. ECCS operation is essential during the LOCA period; therefore, there are no automatic isolation signals.
- gee, valve.
closure requirement W'~ll m hand'cathect by thigh level alarm ln the'ppropr6n4e leandro< B ~llQinq G~rnp, ghiC4 vv>ll 4~ ~~cess>4p- <ncl capite. of p 5 leo.k.a.g< ln+o deco ada rq c ontai ~e~~.
'I P ) h
22.045 You state in 6.2.4.1 of the FSAR that instrumentation lines are designed to the provisions of Regulatory Guide l. 11. Provide the analysis performed which demonstrates that in the event of a rupture of any component in the instrument lines outside the primary containment, the integrity and func-tional performance of the secondary containment and its associated filtra-tion systems are maintained.
RESPONSE
1 An analysis is not warranted since each instrument line that oenetrates the containment includes an excess flow check valve immediately outside the containment as an isolation valve. The leakage from a line ruptured downstream of this valve would therefore be negligible (less than 0.1 gpm) since the check valve will seal the line. The instrument lines from the containment through the excess=flow check valve are Seismic Category 1, meet ASME Section III, Class I requirements if part of the reactor pressure
- boundary, and.have been protected from pipe whip and missiles.
a
. 22.046 Revise Table 6.2-16 of the FSAR to include the isolation provisions for instrumentation lines penetrating the primary containment.
RESPONSE
Table 6.2-16 has been revised to include the instrument lines pene-trating the primary containment (see the response to question 22.027). The individual instrument penetrations fora given type were not listed separately since the information given would be repetitious and can be adequately covered by a single entry. r~vig~,h ) gE l~ C ->- Lb g g bc~a s~4v i9~c( ~.4h g ~ >(ion
g. 022.47 You state in Section 6.2.5.1 of the FSAR that purging hydrogen from the containment is not required as a backup system to the hydrogen recombiners since all components of the recombiners are redundant. We have provided guidance in Section C.4 of Regulatory Guide 1.7 which states that the capability of a controlled purge of the containment atmosphere should be provided to aid in cleanup. Discuss your plans to comply with the guidance contained in Regulatory Guide 1.7. ~Res ense: Refer to 6.2.1.1.8.3, 6.2.5.1.m, and 6.2.5.2.4.*
- See attached draft.
o \\ I
(~ .( << All containment purge valves, including the 2" bypass v.ives, are designed to shut within ou'r seconds of receipt of a containment isolation signal and to shut against full con-tainment design pressure of 45 psig. The contai>>ment 'solat'on s'gnals and the purge valves are part of the co>>tainment isolation system which is an ESF system. Each purge line has two isolation valves. These valves are oponed by allow-ing compressed air to oppose a spring in the valve actuator. On a loss of compressed air, loss of electrical signal, or on a containment isolation signal the valve is shut. If the purge system were operat'ng at the time of a LO A, the system will automatically be secured. The level of the activity released through the purge system before isolation would be limited to the activity present in the coolant prio to the accident since the purge, system will be isolateQ be-fore any postulated fuel failure could occur. 6.2.1.1.8.3 Post LOCA The unit coolers are not, required after a LOCA since heat removal is then accomplished by the containment cooling
- system, a subsystem of the HIIR system, as described in 6.2.2.
~~W~&4eA-.a=. a<,( M .-MC Two 1004 redundant l>ydrogen recombiners are ~~ placed in operation to ensure that the hydrogen buildup does 4 (anirna41E Any equipment located inside the primary containment which is required to operate subsequent to a LOCA has been designed to operate in the worst anticipated acc'de>>t environment
- or the required perioQ of time (See 3.11) 6.2.1.1.9 Post Accident monitoring A description of the post accident monitoring systems is pro-vided, in 7.5.
6.2. 1.2 Containme>>t Subcompartments The two areas within the primary containment considered sub-compartments are the area within the sacrificial shield wall and the area above the refueling bulkhead plate at elevation 583'. All potential pipe breaks within the sacr'ficial shield wall have been evaluated. The information is contained in Refer-ences 3.8-5; 3.8-6 and 3.8-7. These refcre>>ces have been previously submitted to the NRC. Two analyses a"'e being performed to ensure the adequacy of the refueling bulkhead and inner refueling bellows at elevation 583'. The first analysis, a break of the RCIC head spray line, will determine the ma>;imum downward loading due to pipe breaks, and the second analysis, a break of the RRC suction line, will determine the mavimum upward lo ding. These a>>alyses will b-incorporated into the FSAR by means of an amendment. 6.2-33 ~ ~ ~ ~ q ) ~ lf ~ D Q ipr1 D 4 ~ ~ <<<<~+@' P.~V
0 I
k. The system is designed to meet quality assurance, redundancy, power supply and instrumenta"ion re-quirements for an engineered safety feature system. Since the system is redundant and is not shared with other nuclear units, transporta-tion of the hydrogen recombiners is not re-quired. I I -==5 ,/ i 6 ~ 2 ~ 5-2 I, ~c~~~ez" ptlrp~sab'~-ttp-~~~~~ ilot "rv 0<>>Ps: nqa-tYc'thr.i: 'I ~DRj-wgp<~ggc ~f(<f~~~~p~i>>f~~~~~c~~~~A- '.<<...I, 'j...L.,P
System Design
The containment atmosphere control system provides effective control of the hydrogen generated following a postulated LOCA. Piping and instrumentation fo the system is shown in Figures
- 3. 2-17,
- 3. 2-15 and
- 3. 2-6.
Equipment details are given in Table 6.2-17. 'Ene system consists of the following: l. A hydrogen mixing system which operates to assure a well mixed atmosphere in both the drywell and suppression chamber.. This system is the containment spray system and can be ac-tuated approximately 10 minutes after the po tu-lated LOCA. 2. - A hydrogen concentration monitoring syster measures the amount of hydrogen in the drywell and suppression chamber atmosphere. 3. Two 100 percent. capacity hydrogen recombiners, one of which is manually initiated approxi-mately five hours after the accident to preclude the hydrogen concentration from exceeding four percent by volume. The recombiners are cataly-tic type hydrogen oxygen recombiners.
- 6. 2. 5. 2. 1 I<ydrogen Yiixing Sys tern The function of the hydrogen mixing system is to provide a
well mixed atmosphe"e in the drywell and suppression chamber. Since a11 components of the system are redundant, an engineered safety feature containment purge system is not required. However,". containment
- purge, discussed in 6.2. 1. 1.8, has the capabi1ity for
~ a contro11ed purge of the containment atmosphere to aid in hydro- 'en control, if necessary.
0 'e A 1
he cooling water supplied to the aftercooler is returned to the st"ndbv service water system. The cooling water s pplied to the scrubber is discharged to the supp ession pool. components of the containment atmosphere control. system a -e redund nt. Con rois include the control panel located tne main control room and the local contro panel for each ecombiner located in environmentally sui"able rooms the eactor building. All of the functions necessary to con-=ol the system are located in the main cont ol room. 6.2.5.2.4 Containment Purge ).~J w v c 4 r "he-co- "ai n "~"" -y;" - - dukw" eeH.~-..m n-L-purge 6.2.5.3 Design Evaluation Based on the assumptions of the model described below, it 's c lculated that the hydrogen concentration in the cry-we'1 eventually reaches 4S by volume approzimatelv 10.0 hou.-s after the postulated LOCA if the hydrogen recombiner is not in operation. The recombiner is started,
- however,
- shen the hydrogen concentration reaches approximately 3.5" b",
volume (5 hours after the postulated LOCA) to lim't the hvdrogen concentration below 4% by volume. Figure 6.2-26 snows the drywell and suppression chamber hydrogen concen-trat'on as a function of time, with and without operation
- o. the hydrogen recombiner system at design capac'ty of 150 sc m and reduced capacity of 105 scfm.
The determination of the t'me depencent, hydrogen concen-tration in the drywell and suppression chamb r atmospheres is based on a two-region model. of the primary containment, a drywell and a suppression chamber atmosphere; The drywell and supp"ession chamber free volum s contain a'nd water vapo" at atmospheric pressure ju t prior to the postu'ated LOCA. Gases considered available for hvcrogen dilution are the non-condensibles and water vapor present during normal operating conditions. !<ater vapor generated from blowdown is not considered. The radiolytic generation of free oxygen is added to the tota inventory of gases. The pressure in containment is assumed to rema'n at atmospheric pressure and tne temperature h'story of F'g 're 6.2-7 curve a, is used. I '.or'Yawe4~n< gurge, 3<&8'~d 6.2 ~ ~ 4.2,. /. I. 0 h~~5 ~<~~ ~vQYc. '.I't<'76 h +MYAroOBd I ~ ~ p ~r~. nqaroo~n,.oa, a ~ ~ ~ ) g~o q~3 (~i.Fh~:A: P~-)yt)4~'.".+~~ ")0 ~
t.NP-2 Q. 22. 048 You state in 6.2.6.3.1.3 of the CESAR that the co rosion of
- aluminum, zinc, and zinc base paints located either in the drywell or in the suppression chamber were determined to be insignificant.
- However, we have determined that a potential hydrogen release fromm the corrosion of zinc following a postulated loss-of-coolant accident should be considered in the analysis of the total hydrogen production and, accumulation within the containment.
Accordingly, provide the following information: a o Provide the corrosion rate as a function of temperature for all materials in the containment that could become a source of hydrogen due to corrosion. b. Describe how the corrosion rates assumed for the materials identified in Item (a) were established: Identify the experimental data base, including the appropriate references, and-discuss the con-servation in the applicability of the data in view of the calculated environmental conditions following a postulated loss-of-coolant accident. c ~ Provide the mass and surface area of zinc paint and galvanized steel and other corrodible materials in both the drywell and the wetwell. d. Provide a graphic representation of the total hydrogen concentration inside the containment as a function of time with (1) no hydrogen re-combiners operating; (2) one recombiner operating; and (3) both recombiners operating. e. Provide a graphic representation of the con-tribution of each source of hydrogen as a function of-time. Describe the periodic surveillance that will be done to demonstrate the operability of the hydro-gen recombiners and the backup purge system. g, Identify the location of (1) the hydrogen sample points in the drywell and the suppression
- chamber, and (2) the suction and discharge points of the combustible gas control system with respect to nearby structures and equipment.
l'"HP-2 Resnonse: g.'ee the response to question 22.025. The response to the remaining questions will be submitted to tne NRC by March 1979. This delay is due to a recalculation and a review of all.zinc coated areas to confirm our previous statement.
lj ~s ~ MNP-2 22.049 The information presented in Section 6.2.6.1 of the FSAR regarding fluid lines penetrating the containment which will be vented and drained to ensure exposure of the system containment isolation valves to the contain-ment atmosohere and the full differential pressure during the Type A containment integrated leakage rate test, is incomplete. For example, the following systems have not been included in the discussion: (1) the reactor core. isolation cooling system; (2) the reactor pressure vessel instrumentation lines; (3) the neutron monitoring system; (4) the equip-ment and floor drain collection systems; and (5) the primary containment chilled water piping. Identify the status of these lines during the integrated leakage rate test. Provide the following additional information: a. The design provisions that will permit venting and draining of those lines that will be exposed to the containment atmosphere during the integrated leakage rate test. b. Justification for not venting and drainino those eight systems listed on page 6.2-82 of the FSAR. E
RESPONSE
A tabulation of isolation valve status during Type A tests is given below. Included is the justification for not venting systems as applicable. Also note that l<NP-2 does not have a primary containment chilled water system but rather only the reactor closed cooling system that penetrates contain-ment. ~Sstem
- 1. Reactor Closed Cooling 2.
RHR 3. HPCS CTHT ISO Valve Status - T e A Test Open Closed (except pump min. flow), if required to cool reactor, necessary valves will be open. Cl os ed Justification For Not Ventino a. Required to main-tain CTHT tempera-ture durina ILRT a. System is normally filled*with water and is operated under post-LOCA conditions.
- b. Required to cool reactor during shutdown.
a. System is normally filled*with water and operated under post-LOCA conditions'
lNP-2 ~Stem 4. LPCS
- 5. Standby Liquid Control
- 6. Reactor Water Cleanup 7.
Feedwater System CTNT ISO Valve Status - T e A Test Closed (except suction line) Closed Closed (except suction line at feedwater) Closed (except for N.O. gate) a ~ a ~ a. b. 'a ~ b. Justification For Hot Ventin System is normally filled*with water and operated under post-LOCA conditions. System is normally filled. Required to be filled during all phases of plant operations System is normally filled with water. System suction is off the RRC pump suction line which must be flooded to maintain reactor water level during the Type A test. Consequently, the suction piping will remain flooded also. System is filled with water. System will poten-tially operate post-LOCA. 8. RCIC 9. Equipment and Floor Drain Systems Closed (Steam supply to iso-lation valve drained and vented. Down-. stream of isolation valves also vented.) Closed a 0 b. a. b. System is normally filled*with water. System will operate post-LOCA. System will be water sealed at its low point. During and following a LOCA the sumps will be filled with water and thus seal these systems from the con-tainment atmosphere.
WNP-2 ~Setem 10. Reactor Recirc. Sampling CTMT ISO Valve Status - T e A Test Closed Justification For Not Ventin a. System wi 1 1 be filled with water. b. System ties into the RRC system which must remain full during Type A test and conse-quently the sampling system remains filled also.
- 11. Control Rod Drive Line up oer scram configuration a
~ The seismic portion of the system is water sealed. The nonseismic portion of the system is vented.
- 12. Neutron Monitoring 6all Valves Closed System (TIP) 13.
Reactor Instru-mentation Lines a ~ a. The TIP mechanism will be withdrawn and the ball valves closed. The TIP system will be exposed to the Tyoe A test pressure to the extent of system leakage. The RPV wil be vented as will the related instrumentation. This instrumentation will thus be exposed to the Type A test pressure.
- Water leg pumps provided to maintain water level in the pipe.
All piping to be exposed to the Type A test pressure is provided with low point drains and vents to ensure adequate removal of water from the effected systems. The RCIC steam supply and the reactor instrumentation will then be exposed to the Type A test pressure since the reactor vessel is vented.
llNP-2 g,'33 '3 I 22.050 Those closed,.~y~feyst o~itside, containment, which must function following an accident be'come'xtensions of the containment boundary after a pos-tulated loss-of-coolant accident. Certain of these systems may also be identified as one of the redundant containment isolation barriers. Since these systems may circulate water or the containment atmosphere which may circulate water or the containment atmosphere which may be contaminated, components of these particular systems which may leak are relied on to provide containment integrity. Accordingly, provide the proposed leakage limit for each system that becomes an extension of the containment boundary following a postulated loss-of-coolant accident. Discuss your plans for leak testing the systems either hydro-statically or pneumatically. Discuss how the process leakage limits will be included in the radiological assessment of the site.
RESPONSE
Closed systems which become an extension of the primary containment during the post-LOCA period are the: High Pressure Core Spray (HPCS) Low Pressure Core Spray (LPCS) Residual Heat Removal (RHR) Containment Atmosphere Control (CAC) Reactor Core Isolation Coolant (RCIC)* The above systems will be given a hydrostatic or pneumatic test following system completion per Section III of the ASME Code. This test will be repeated once every ten years per the requirements of Section XI of the ASME Code. In addition to these tests, the
- HPCS, LPCS, RHR and RCIC (water side) systems are maintained under constant pressure.
These systems are kept full of water and pressurized by either water leg pumps or by the suppression pool static head. During normal operation any sig-nificant leakage from valve packing or pump seals in these systems
- would, be detectable by. visual observation or by sump level alarms.
Also the pumps will be flow tested quarterly and the valves will be exercised per the ASME, Section XI inspection program. During the tests a visual inspection for water leakage from the component will be made. In addition, All containment isolation valves in the above systems subject to reactor .. P,..!-.3P-,,'.,.~! i 3 P f 3 tg 3 tii "f ii iit" 3"""i:3 i Pt, f Pi 3. -3 (Note 5), 7, and 8 (Note 4)).
- The RCIC system is included here as it is a closed system outside contain-ment and may briefly operate following a LOCA. It is not required to operate post-LOCA, however, and will be isolated once reactor pressure is reduced.
MNP-2 Defining the threshold value at which leakage from a component would become visible is.~iffjcvl,+,,however the value canegal,,istically be assumed to be below .1 gpm. Based on this and the fact that excess leakage paths would be repaired, any water leakage during the post-LOCA period is expected to be negligble. The RCIC steam lines from the reactor to the inside containment iso-lation valve and the CAC system will be open to the primary contain-ment atmosphere during the Type A test. Leakage from these lines will be part of the measured allowable leakage for the Type A test. Al.l these. svstems are entirel.v within.the..secondarv containment ancl conse-. quently any leakage during the post-LOCA period will be processed by the standby gas treatment system (SGTS). Appendix B to the Standard Review Plan 15.6.5 deals with leakage from ESF systems during a LOCA and states "when ESF - grade filters, are supplied, no doses resulting from passive failures need be considered." Sihce the SGTS includes ESF - grade filters these doses have not been included in the site radiological assessment total LOCA dose figure. However, an analysis has been done to confirm the fact that the doses resulting from passive failures such as minor leaks in ESF systems is negligible and is outlined in the response to question 312. 12.
I I-
22.051 Discuss your plans for including in the Type A tests, the reactor building pressure sensing lines which will become extensions of the containment boundary following a postulated loss-of-coolant accident. RESPONSE:. The reactor building pressure sensing lines will be valved off at the instrument isolation valves.. This is done to preclude instrumentation damage upon exposure to the high pressure associated with Type A tests. Also this. prevents the ECCS related pressure switches from activating with the. high drywell pressure of the Type A test. The Type A test will include all tubing up to the isolation valves however. No Type C tests are planned with these lines either as they are designed to the requirements of Reg. Guide 1.11.
22.052 Identify the valves for which the test pressure is not applied in the same direction of the pressure ex'isting when the valve is required to perform a safety function. Provide a demonstration that the valve leakage rates for these particular valves are equivalent to or greater than the leakage rates whi.ch would occur if the test pressure were to be applied in the same direction as the pressure existing when the valve is required for its safety function.
RESPONSE
The response to question 22.010 identifies the containment isolation valves that will be tested in the reverse direction from their safety function. The justification for this is also given in that response. Note also that the gate valves have a shop leak test applied to both sides of the valve. These test results have been reviewed and found acceptable. The results for these tests are on file at the MNP-2 site and are available for inspection.
Cr gC S 1 i pll C 'tg IP ilp(~ k}}