ML20213D345
| ML20213D345 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 09/26/1980 |
| From: | Check P Office of Nuclear Reactor Regulation |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-0314, CON-WNP-314 NUDOCS 8010230566 | |
| Download: ML20213D345 (26) | |
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o NEMORAtlDUM FOR:
R. L. Tedesco, Assistant Director for Licensing l
FROM:
Paul S. Check, Assistant Director for Plant Systems, DSI
SUBJECT:
REVIEW 0F WNP ADDITI0fuu. QUESTI0fiS Plant Name:
WPPSS fluclear Project flo. 2 Docket flo.:
50-397 l
Milestone flo.:
05-21 Licensing Stage:
OL Responsible Br. and LPft:
LB-1/M. D. Lynch s
Systems Integration Branch Involved:
RSB Review Status:
Incomplete Enclosed are the additional questions prepared by Savannah River Plant personnel working in support of the RSB review of the UflP-2 FSAR. This constitutes those questions generated in the laboratory review through A'nendment 9.
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B. Youngblood ff. Lynch T. Speis T. Collins G. Mazetis W. Hodges C0flTACT:
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Docket File V SEP 2 61910 RSB R/F i
PCheck MEMORANDUM FOR:
R. L. Tedesco, Assistant Director for Licensing i
j FROM:
Paul S. Check, Assistant Director for Plant Systems, DSI '
SUBJECT:
REVIEW 0F WNP ADDITIONAL QUESTIONS Plant Name:
WPPSS Nuclear Project No. 2 Docket No.:
50-397 l
Milestone No.:
05-21 Licensing Stage:
OL Responsible Br. and. LPM:
LB-1/M. D. Lynch l
Systems Integration Branch Involved:
RSB i
Review Status:
Incomplete i
Enclosed are the additional questions prepared by Savannah River Plant i
I personnel working in support of the RSB review of the WNP-2 FSAR. This
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Ar.endment 9.
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SEP 2s 2L"u MEMORANDUM FOR:
R. L. Tedesco, Assistant Director for Licensing FROM:
Paul S. Check, Assistant Director for Plant Systems, DSI
SUBJECT:
REVIEW 0F WNP ADDITIONAL QUESTIONS Plant Name:
WPPSS Nuclear Project No. 2 Docket No.:
50-397 Milestone No.:
05-21 Licensing Stage:
OL Responsible Br. and LPM:
LB-1/M. D. Lynch Systems Integration Branch Involved:
RSB Review Status:
Incomplete Enclosed are the additional questions prepared by Savannah River Plant personnel working in support of the RSB review of the WNP-2 FSAR. This constitutes those questions generated in the laboratory review through Amendment 9.
i pk, g aul S. Check, Assistant Director v
for Plant Systems Division of Systems Integration cc:
B. Youngblood M. Lynch T. Speis T. Collins G. Mazetis W. Hodges CONTACT:
T. Collins, NRR X-27803
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REACTOR SYSTEMS BRANCH 0211.107 Regulatory Guide 1.70, Revision 3, Section 3.5.1.2, requires that (3. 5.1.2) the structures, systetts, and ccmponents protected by physical barriers should be identified. The discussicn and the figures in the FSAR do not indicate where, if at all, physical missile barriers are used.
d Identify all structures, systerns, und cceiponents 'that *:re protected by physical barriers.
Provide a descriotion of the types of physical barriers that are ecployed at your plant, Q211.108 Section 3.5.1.1.2 of the FSAR states that missile trajectcries (3.5.1.2)
-are selected to encompass the most adverse cond2tiens. It is r.ot clear fran the infonnation prcvided ir, the F3AR what the trajectories of the credible primary missiles would be and what systems might be disabled by the missiles.
Provide the bases for selection of the probable missile trajectories and show the trajectories on the appropriate FSAR figtre. Incitde a discussion on the system, component, cr structure that could be damaged or disabled by a missile. The extent of damage from each missile should be discussed.
Q211.109 Section 3.5.1.132 states that thennowells and sample prebes (3.5.1.2) do not present potential hazards es postulated missiles affecting safe shutdown.
Provide justification to support this position en the thermowells and semple probes.
Q211.110 The notations, "251 BWR/5-MSIV,14d/5 Void Coefficient" on Figures (5.2.2) 5.2-4 and 5.2-5 indicate that these figures may be generic and not specifically for WNP-2.
Confirm that these figures are applicable to WNP-2.
If these curves are not applicable to WNP-2, cceplete the necessary analyses to provide data similar to that. new presented on Figteres 5.2-4 and 5.2-5.
l 0211.111 Article NB-7200, Overpressure Protection, of the ASME Boiler and-(5.2.2)
Pressure Vessel Code,Section III, requires that an overpressure l
protection report be provided. No overpressure report could be i
found in the FSAR. Provide this report.
1 Q211.112 Section 5.2.2.4.2.1 of the FSAR states that cyclic testing has (5.2.2) demonstrated that the safety / relief valves are capable of at least 60 actuation cycles between required maintenance. Will the actuations of the safety / relief valves recorded? If so, how will I
these data be recorded and reported to the NRC7
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4 0211.113 It would appear that improper setpoints would be a credible ecmon (5.2.2) mode failtre Wich could result in degradation of the pressure relief systems. Show that adequate safety margin has been incitded in the overpressurization analysis to protect against a ccanon : code failure of the safety / relief valves to open at the prescribed values.
0211.114 Subsection 5.2.2.4.1 of the FSAR states that each safety / relief (5.2.2) valve is provided with a device to counteract the effects of backpressure Wich results in the discharge line when the valve is open and discharging steau. idhat type of device is provided?
Cescribe the device and what effects would be anticipated if the device were to fail.
0211.115 subsection 5.2.2.4.1 of the FSAR states that setpoints for the (5.2.2) ocwer actuated mode for each safety / relief valve are specified in Table 5.2-2.
Table 5.2-2 provides a listing of the setpoints and velve cacacities of the valves in the five safety mode groups (spring-operated mode), but no data are presented for the relief mcde of operation.
Provide the relief setooint for each safety / relief valve in Table 5.2-2 and in Figure 5.2-6.
4211.11c Previde the results of hydraulic calculations that show the Mach (5.2.2) ntaber, pressure, and temoerature at various locations fece upstrean of the safety / relief valves to the suppression pcol at maximta flow conditions. 'Ihe concern is related to the potential fcr the develognent of damaging shock waves to the discharge piping. Incitde the effects cf suppression pool swell variations on the operation of the safety / relief valves.
C211,117 Resolve the following inconsistencies:
(5.2.2) a) Figure 3.2-2 of the FSAR indicates in details B and C that the instrtment air supply lines to the safety / relief valve air acetmulators are safety class G (non-safety grade). Figure 9.3-2 shows these lines as safety class 2 cr 3 (safety grade).
b) Figure 5.2-5 shows the safety / relief valves assigned to the automatic depressurization function are F013-M, -N, -P, -R,
-S, -U, and -V.
Figure 9 3-2 shows the dual acetmulators used for the ADS valves assigned to safety / relief valves F013-D,
-E, -H,
-J, -M,
-F, and -S.
Q211.118 Subsection 5.2.2.4.1 cf the FSAR states that the pnetmatic (5.2.2) scotuulater provided for each safety / relief valve has sufficient capacity to p ovide one safety / relief valve actuation. Figure 3 2-2 irdicates that the air supply lir.; W;trem of the ball check valve is safety class G (non-safety grade). If the air line were to break upstream of the ball check valve, would there be an ir.dication in the control room of this break and an indication of the sectnulator status? If an indicaticn is given, what operater a::tibn would be required? Also, show that acetmulator capacity fer cne sctuation is sufficient.
Q211.119 Subsection 5.2.5.2 of the FSAR indicates that temperature and (5.2.5) pressure monitoring devices are used as primary detection devices for midentified leakage. Regulatory Guide 1.45 states that hunidity, temperature, or pressure mcnitoring should be considered as alarms or indirect indications of leakage. Justify this exception to the criteria of Regulatory Guide 1.45.
Canonstrate that the midentified leak detection systems can detect leakage on the order of one gallon per minute in a one-hour period.
Q211.120 Subsection 7.6. 1. 13 7 of the FSAR states that the same leak (5.2.5) detection monitor (a three-channel unit) will detect both airborne particulate and gaseous activities in the drywell atmosphere using seintillation detect crs.
Explain tow these two different types of airborne activities are separated by the monitor.
Justify taking credit for both monitoring technigt.es in subsection 7.6.2.4.2.1.2 while using the same device. State the sensitivity and response time of the radioactivity monitor.
Q211.121 Subsection 5.2.5.5.5 of the FSAR states that the leak detection (5.2.5) system will satisfactorily detect unidentified leakage of 5 gpn.
Subsection 7.6.2.4.2.1.2 states that the sensitivity and response time for each portion of the leak detection system for detection of unidentified leakage is one gallon per minute in less than one hour (excluding airborne systems).
Resolve this inconsistency.
Q211.122 The response to Q211.007 requires additional information. It is (5.2.5) unclear how the ecmparison will be made between the radioactivity monitoring and the sunp level monitoring.
Cescribe briefly the mechanics of making these data ccmparisons.
What calibration and operability verifiestion tests will be perfonned for each independent leakage detection system? Which leakage detection system is to be used as the reference for comparison with the other systems? Do the radiation monitoring systems have radioactive sources (check sources) built into the systems?
0211.123 Identified leakage is determined during pre-operational testing or (5.2.5) is measurable during reactor operation. Will the identified leakage be measured regularly and recorded? If so, provide the frequency that these data will be recorded and indicate what procedural guidelines are to be used to change the magnitude of the base identified leakage rate?
0211.124 It is unclear in subsection 5.2.5.2f of the FSAR whether (5.2.5) ccmparative "gr ab" samples of the continuously monitored containment atmosphere can and will be taken on a periodic basis.
Resolve this ambiguity.
If " grab" samples are not to be taken, justify the cmission of these ecmparative data.
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,3 Q211.125 Standtrd Review Plan 5.2.5 specifies that unidentified leakage (5.2.5) should be collected separately frcm the identified leakage so that a mall unacceptable unidentified leak is not masked by larger acceptable identified leakage. Section 5.2.5 of the FSAR does not clearly indicate that s+parate collection of identified and midentified leakage is provided.
Provide assurancas that identified and unidentified leakage will be collected separately. If separate collection is not to be provided, provide justification for use of a ecmon collection reservoir and show that a mall taidentified leak of about 1 gpu would be reccgnized within one trur.
Q211.126 Provide a list of all indications available to the control roca (5.2.5) operator for evaluating and detecting unidentified leskege. Show how the operator will determine the amotnt of leaktge by observing the indications that are available to him, incitding the need for unit conversion (cotnt rate to spn, etc). If the conitoring is ccmputerized, discuss the backup procedures available should the ccmputer beccme inoperative.
Q211.127 Resolve the following inconsistencies in Figure 5.4-14b:
(5.4.7) a) In mode B of the RHR system operation, there is an (nexplained 500 gpn increase in flow in goirg; frca process points ISB to points 21B and 23B.
b) In mode C-1, the stm of the flows past process points 40 and 40.2 should be equal to the flow past point 19 As presented, the sta of flows past points 40 and 40.2 is twice the flow that is tabulated for point 19 c) In mode E, it is not clear -hat the total system flow shculd be (14,900 or 7450 gpn).
d) In the stmary of the various modes of RER system operation, reference is made in mcde D to note 13 Note 13 has been deleted from the P&ID. Provide suppleental information to mcke this reference meaningful or delete this reference altogether.
e) Subsection 5. 4. 7.1.1. 2 and Table 6.3-2 of the FSAR state that r
the functional design bases fcr the LFCI mode of RHR operation is to ptmp 7067 gpn of water per loop into the reactor core region of the reactor vessel. Figure 5.4-14b and the response to Q211.038 state that each loop should supply 7450 gpn to the reactor core region under accident conditiens.
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7 0211.128 Subsecticn 5.4.7.1 of the FSAR states that spoolpiece interties (5.4.7) are provided to pemit the RHR heat exchangers to be used to supclement the fuel pool cooling system.
Describe the administrative ccntrols that will be exercised for the tre of these spoolpieces. khat '.culd be the effects if the spoolpieces were left in place and the RHR system were operated in any or all of the RHR modes of operation? Similarly, a spoolpiece is showa on drawing N521 that connects the low pressure core spray (LPCS) system to the EHR loop A suction pipe. Describe the ptrpose of this intertie and, also, describe the effects on both the L.PCS and EHR systems if the spoolpiece were inadvertently left in place. Are the same administrative controls used for the fuel pool ecoling system spoolpiece used for the LPCS spoolpiece?
Q211.129 The standby liquid centrol system and the recirculation flow (4.6) control systsn are reactivity control systems. Address or reference these systens in Section 4.6 and address all requirements of Standard Review Plan 4.6.
C211.130 Table 13-5 Indicates scecific cesign changes frcrn the PSAR (4. 6.1.1. 2. 4 ) to the FSAR for the CR0 system. 'Ihe design changes for the CRD return line modification Meressed in Question 211.19, have not been incitried in the text description of the FSAR and Figures 4.6-Se, 4.6-5b, and 4.6-6a have not been revised.
Revise the text description in the FSAR to reflect the specific design changes in Table 1.3-8 for the CRD syste and ecdify the above figures acccrdingly.
Q211.131 The scram discharge voltme header piping is sized to receive (4.6.1.
and contain all water disharged by the centrol red drives during 1.2.4.1) a scram, independent of the instrtmr.nt volune. Show quantitatively how a minimun volute cf 3 34 gallons per drive is required since approx 11cately 4 gpt is required to insert the rods with uo to an additional 0.34 gun required for coolir4 C211.132 Resolve the folleving items relating to filtration of (4. 6. L 1.
ccndensate water for the CRD hydraulic system.
2.4.2.1) a) The text description indicates that r.ormal filtration of condensste water on the suction side of the CRD water ptrap is accomplished by a 25 micrcn filter cnd that a 250 micron strainer is providW in the bypass line for the 25 micrcn filter. hen it is being serviced. Figtre 4.6-6a indicates that double t'iltration of condensate water on the suction side of the CRD water ptrap nomally ocetrs via a 250 micron strainer aryd a 25 micron filter in series. F.xplain this discrepancy.
b) Describe provisions in the W-2 design to crotect the hydraulic control units (.uct s) and control rod drives (CP.Es) frcm dafntge dtg to inadvertent failtre of either the ptn:p stzetion filter er the drive water filter. If none exist, provide justification that inadvertent failure of either filter will r.ot cause dirtage to the HCUs and CRCs.
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C211.133 In Figtre 4.6-5b and Drawing M528, pressure tranmitter (N005)
(4.6.1.1.
transits a signal to a pressure switch (N6CO) in the process 2.4.2.2) instrtmentation panel in the control rocm, which energizes an annunciator in the control rocm at any time pressure in the charging header falls below the setpoint. Explain why an alarm on high is indicated for the pressure switch (N600) instead of an alarm on low eich would provide protection against charging header pressure falling below the setpoint.
C211.134 In the text for the CRD cooling water header, there is no (4.6.1.1.
discussion of valves F129, F130, F131, and F132 which are shown on 2.4.2.4)
Figtre 4.6-5B.
'Ibese valves are not incitded on Drawing M528.
Explain this discrepancy and update the FSAR accordingly.
0211. 135
'Ihe text description of the scram acctmulator indicates that (4.6.1.1.
a check valve in the acetmulator charging line prevents loss 2.4 3.9) of water pressure in the event supply pressure is lost. 'Ibe symbol for valve 111 in Figure 4.6-5b and Drawing M528 appears to I
be that of a normally open globe valve instead of a stop-check l
globe valve. Explain this apparent discrepancy.
l C211.136 Identify the specific cccmon mode failure analysis and protection (4.6.2) frcm comon mode failures referenced in Section 15A by Sections 4.6.2.1 and 4.6.2.2, respectively.
0211.137 Identify the layout studies done to assure that no interference (4.6.2 3 1.2) exists dich will restrict the passage of control rods and the pneoperational test (s) that are used to show acceptable l
performance.
C211.138 Provide the comon mode failtre probability value for the control (4.6.4.1) rod drive and the standby licuid control systems.
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G Q211.139 Provide the following information concerning RCIC equipment and (5.4.6.2.2) ccmponent descriptions.
a) Section 5.4.6.2 of Regulatory Guide 1.70 states that significant design paraneters for all components of the RCIC system be identified and that all components be shown on appropriate P&I diagrams. Lesign parameters for only a portion of the RCIC components are incitded in Section 5.4.6.2.2.2.
Scme of the more knportant ecmponents cmitted are the:
- 1) Water leg ptanp
- 2) Barcmetric condenser
- 3) Vacuun tank
- 4) Condensate punp
- 5) Turbine and steam supply drain pots
- 6) Turbine governing and trip throttle valves
?) Pump suction strainers in the suppression pool Provide the significant design parameters for all RCIC com;cnents not included already in Section 5.4.6.2.2.2 and verify that each component can be identified on Figures 5.4-9a and 5.4-9b.
b) The RCIC turbine is identified as component C001 in Section 5.4.6.2.2.2 and as ecmponent C002 in Figure 5.4-9b.
Correct this discrepancy.
Q211.140 Four keylocked valves (F063, F064, F068, and F069) are indicated (5. 4. 6. 2.1. 3 ) in step "a" as electrical interlocks. However, one of these valves, valve F064, is not indicated as keylocked in Figure 5.4-9a, while valve F008 is indicated as keylocked.
Resolve this discrepancy.
Q211.141 Is the RCIC electro-hydraulic system integrated with the turbine (5.4.6) governing valve of a safety grade design (i.e., Seimic Category I)?
Q211.142 Describe the design features and operating procedures that (5.4.6) preclude water hamner effects at the punp discharge of the RCIC system.
Q211.143 Show how the gre-operational initial startup test progrsms for the
]
(5.4.6.4)
RCIC system in Section 14.2.12.1.8 meet the intent of applicable i
sections in Regulatory Guide 1.68.
Q211.144 The ASiE Boiler and Pressure Vessel Codes,Section III, Article (5.4.6)
NB-7000 requires that individual pressure relief devices be installed to protect lines and ecmponents that can be isolated from nonnal system overpressurization protection. With reference to appmpriate P&ID, identify those pcrtions of the RCIC system that can be isolated frcm normal system overpressure protection.
Discuss the relief devices provided or provide the basis for deciding that relief devices are not required.
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l Q211.145 At scue BWR installations, the check valves in the turbine (5.4.6) exhaust line of the RCIC system which serve a containment isolation function have been damaged as the result of intemittent j
closure. The intermittent closures arise frcm flow escillaticns in the exhaust line associated with romation and collapse of j
steam bubbles in the suppression pool. One type of corrective action involves use of a sparger on the exhaust piping in the suppression pool to reduce the flow oscillations.
a) Is the 10" exhaust pipe shown in Figure 5.4-9a installed as a sparger for this purpose?
b) Are there other design features used at WNP-2 to prevent this type of damage?
C211.146 In the responses to Questions 211.046 and 031.015, it is stated i
(5.4.6) that an autcmatic safety-grade switchover frcm the condensate storage tank to a Seismic Category I supply (i.e., the suppression pool) has been provided as a convenience to the operator.
Provide a description of the automatic switchover feature and its initiating signal and confim that both electrical and mechanical features are safety grade.
Q211.147 The text indicates all components of the RCIC systen are capable (5. 4. 6.1. 2.1 ) of individual ft:nctional testing during normal plant operation.
(5. 4. 6. 2. 4 e) Table 13-8 indicates each component, except the flow controller, is capable of functional testing. Resolve the discrepancy with respect to functional testing of the RCIC flow controller.
Q211.148 Resolve the following items in Table 15.0-2:
(15.0) a) Modify the values of vessel level trip to agree with the values specified in Figures 5.2-6 and 5 3-2 (item 29).
b) Specify the maximtm percent relieving capacity asstmed in Chapter 15 for each mode of SRV actuation (items 25 and 26).
c) Provide the following infomation concerning the high flux trip setpoint used as input to the REDY model (item 29):
- 1) Explain why the high fita trip setpoint should not be increased to 122% NBR prior to multiplication by the themal-power correction factor of 1.043 to account for the setpoint plus calibration error, instrtment accuracy, and transient overshoot specified in Table 7.2-4.
- 2) Explain 41y the thermal-power correction factor is applied to the high fits trip setpoint used in the REDY model.
d) Provide the following information concerning the APRM thermal trip setpoint used as input to the REDY model.(item 30):
- 1) Specify the highest flow-related trip setpoint to be given in the Technical Specifications and how this value is obtained.
- 2) Is the 122.03% NBR setpoint equal to the setpoint to be specified in step d)1) times the themal-power factor of 1.043 specified in step c)1)?
e) Table 15.0-2 does not contain all of the input paraneters used in the REDY ccmputer code. For each transient and accident analyzed in Chapter 15, provide the following:
- 1) A list of all input parameters.
- 2) Justificatien that the input parameters are conservative.
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0211.149 Provide a realistic range and permitted operating band for tne (15.0) exposure dependent parameters in Tables 4.4-1 and 15.0-2.
In Table 15.0-2, provide assurance that values of parameters selected yield the most conservative results.
0211.150 Provide a listing of the transients and accidents in Chapter 15 (15.0) for which operator action is required in order to mitigate the consequences. For corrective activas required prior to 20 minutes, provide juttification.
0211.151 The analysis of transients and accidents in Chapter 15.0 does not (15.0) state which of the RPS time response delays in Table 7.2-5 is used in the REDY computer model (NED0-10802). For each transient and accident in Chapter 15.0, specify which delay time in Table 7.2-5 is used in the analysis and thy the specified delay time is conservative.
Q211.152 In relation to Figure 15.0-2, confirm the following items for all (15.0) transients in Chapter 15.0 which require control rod insertion to prevent or lessen plant damage:
a) The scram curve used in Chapter 15.0 analyses (Figure 15.0-2) has a total reactivity worth of $37.05 and is the neminal scram curve multiplied by the standard transient safety conservatisn factor of 0.80.
b) The slowest allowable scram insertion speed was used for the scram curve applied to Chapter 15.0 analyses.
cl The end of cycle 1 scram curve has a, total reactivity worth of
$40.21 and is identified incorrectly in Figure 15.0-2.
Q211.153 For transient analysis, credit has been taken for safety / relief (15.0) valve (SRV) actuation only in the relief mode. A more conservative approach would be to take credit for SRV actuation in the safety mode, resulting in higher peak vessel pressures.
a) inhat quantitative effect on MCPR and peak vessel pressure does credit for SRV actuation only in the safety mode have on each transient analyzed in Chapter 15?
b) In Section 5.2.2, the relief mode appears to be nonsafety grade because credit for 50% relieving capacity associated with power-actuated pressure-relief valves in ASME B&PV Code Section III, NB-7000, was not assuned for overpressure protection. Are all equiptent and components required for SRV actuation in the relief mode nonsafety grade?
If not, identify specific equipnent and ccmponents that are safety grade and those which are nonsafety grade, c) If the relief mode is ncnsafety grade, explain why credit was taken for this mode of SRV actuation in Chapter 15.0.
If the relief mode is safety grade, explain why credit for SRV actuation with up to 50% relieving capacity in the relief mode and additional relieving capacity up to 50% as required in the safety mode was not applied to analyses in Section 5.2.2 and Chapter 15.
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0211.154 Modify the sequence of events tables in Section 15.0 to specify (15.0) the opening anc closing times of referenced valves and the time at which each reactor vessel alarm or trip water level is attained throtshout the duration of each transient in places where this information is not already incitded. Incitde appropriate delay times frcm the initiating signals and confinn the delay times are applied consistently between the event tables.
l i
Q211.155 Modification of NSOA drawings to incitde use of nonsafety-grade (15.0) systems or components Mitch mitigate transients and accidents was requested in Question 211.85. In conjunction with this request:
)
a) Provide a table of the nonsafety-grade equipnent and i
cceponents asstmed to mitigate consequences for each.
b) Provide the A(ACPR) and A(Apeak vessel pressure) that would result if only safety-grade systems or ccmponents were asstmed in the analysis for each event in Section 15.0 that takes credit for specific nonsafety-grade systems or components.
0211.156 Discuss how the pre-operational and startup tests will be used to (15.0) confirm flow paraneters used in Chapter 15 analyses.
Provide details of any previous test of components in test facilities conducted to show satisfactory performance of the recirculation and feedwater flow control systems and respective ptmps.
Q211.157 Analyze the turbine trip and generator load rejection transient (15.0) from a safe shutdown earthquake event. Credit should not be taken for non-seisnically qualified equipnent or any equipnent contained in a non-seismic structure.
0211.158 Q1 page 4-7 of NEIO-10802, it is stated that the difference in (15.0) trend of flow coastdown versus initial power between the analytical and experimental coastdown curves for Dresden Unit No.
2 (a EWR/3) in Figtre 4-11 was due in part to differences between actual and ccmputed jet punp efficiencies, a) How has this effect been treated in analysis of kWP-2 transients-involving flow coastdown with two recireteation ptmp trips?
b) Is this treatment applicable to kHP-2 which is a BWR/57
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0211.159 GE calculations perfomed for decrease in reactor coolant (15.0) temperature (Section 15.1) and for reactor pressure increase (Section 15.2) events using the proposed ODYN licensing basis model (NEDo-24154) have shown that in some cases a more limiting CPR is predicted than by the current REDY licensing bases model j
(NEDO-10802). Since Question 211.49 was submitted, the CDYN model has been approved. Based on a letter to Glen G. Sherwood dated 1/23/80 from Richard P. Denise, the staff's CDYN licensing position is that GE can proceed with ODYN analysis of certain events described in Section 15 of licensing application Safety Analysis Reports.
Provide the following additional information in conjunction with Question 211.49:
a) An ODYN analysis of the applicable events (One-D) listed in Tables 2-1 and 2-2 of NEDE-25154-P.
b) A list of all input parameters for each event.
c) Justification that input parameters for each event are 1
conservative.
Q211.160 For each transient and accident analyzed in Section 15, identify (15.0) each nomally operating system for which credit has been taken.
Q211.161 Provide assurance that the limiting puno trip is assuned in (15.0) analyzing decrease in reactor coolant system flow rate transients.
Different trip signals may cause different coastdown characteristics.
Identify the trip signal that can be expected to produce the most severe punp coastdown characteristics.
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Q211.162 In the analyses for the generator load rejection and turbine trip (15.0) transients, credit is taken for imediate reactor scram and recirculation punp trip obtained from a valve closure signal (turbine control valve for load rejection and turbine stop valve i
for turbine trip). Analyze these transients without taking credit for imediate reactor scram and recirculation punp trip. Take credit only for safety-grade, seisnic Category I equipnent and assune loss of offsite power. What is the effect of the failure of a single safety-grade ccmponent? Provide the effect on analytical results that WNP-2 operation with the new 8 x 8 fuel design with two water rods will have.-
Present curves similar to those of Figures 15.2-2 and 15.2-4 and give values of maximun vessel pressure and minimun MCPR with the times at which these values occur. Evaluate the percent of fuel rods which would reach boiling transition. Since this event is not an anticipated transient, limited fuel failure can be allowed if dose consequences are acceptable.
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Q211.163 For the majcrity of events analyzed in Chapter 15, the (15.0) recirculation flow control mode (autcmatic or manual) asstmed in the analysis is not specified. Our concern is that the mode selected may not result in the most severe margins on MCPR and peak vessel pressure.
a) Soecify the recirculation flow control mode asstmed for each event analyzed in Chapter 15.
b) Specify the change in MCPR and peak vessel pressure that results in these paraneters for each event if the opposite recirculation flow control mode had been asstned in the analysis.
C211.164 On page 15.1-2, it is stated that the thermal power monitor (TPM)
( 15.1.1. 2.2) is the primary protection system for mitigating the consequences of the transient resulting from loss of feedater heating. A description of this monitor, which typically involves the flow-weighted APRM scram in conjunction with a 6-second time constant circuit, was not found in the WNP-2 FSAR. Provide this this description in sufficient detail to permit evaluation of the TPM for WNP-2.
If the time constant, which affects scram initiation by the TPM, is less than the effective time constant for the WNP-2 fuel for this type of transient, the TPM should provide a conservative measure of the time variation in surface heat flux. However, if the time constant is appreciably larger than that for the fuel, the fixed APRM trip without a time constant would provide the scram grotection. The resulting MCPR would then be less than that predicted for the TPM scram which has a lower setpoint.
There is no etrrent provision in the Technical Specifications for surveillance of this time constant circuit.
It is the staff's position that credit be taken only for the fixed APRM scram in Chapter 15 unless the TPM is approved by the staff and appropriate limiting conditions for operation and surveillance requirements are incorporated in the Technical Specifications for WNP-2.
a) Provide an analysis of the " loss of feedwater heating" transient asstming credit only for the fixed APRM scram. This is a more conservative approach because it will result in a more severe transient due to the higher fixed APRM scram setpoint.
b) Revise NSCA Figure 15A.6-21 to indicate the high flux scram signal occurs frcm the fixed APRM scram instead of the TPM.
c) Re-evaluate single failure criteria in Section 15.1.1.2.3 without taking credit for the TPM.
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a) Explain the discrepancy between the asstmed feedwater controller failure valves at maximun demand specif.ied in tne text (135% flow) and in Table 15.1-3 (146% flow). Provide the basis for selecting the magnitude of W flow increase asstmed in the analysis, b) In conjtnetion with the magnitude of feedwater (W) increase asstmed in the analysis, explain why the full W increase is attainea at essentially zero seconds in Figure 15.1-3 In GESSAR 238-732, the W increase is initiated at zero seconds and attains the full value (maximun demand) at approximate 3y 5 seconds.
i c) If the W temperature at the reactor vessel has been asstmed constant, provide a quantitative analysis that includes the effect of FW temperature variation on MCPR and the basis for determining this variation.
Incorporate any changes frcm step a) above concerning the appropriate value of W flow rate asstmed in the analysis of this transient.
0211.166 The gressure regulator failure at 115% NBR steam flow is simulated (15.1.3.3 2) in Figure 15.1-4 in a manner consistent with GESSAR 238-732.
However, the asstmed pressure regulator failtre value of 115% NBR stean flow for WNP-2 appears low ccmpared to a failure value of 130% steam flow used in other FSARs with approximately 15% greater than the nonnal maximun flow permitted by the stean flow limiter.
a) Explain the difference between the 110% NBR steam flow indicated as the nonnal maximun flow limit in this section and the 115% value specified in Section 15. 1 3. 1. 1.
b) Explain the basis for selecting the asstmed pr ssure regulator failtre value of 115% NBR steam flow used in the FSAR.
If a new steau flow value in excess of that permitted by the steam flow limiter is chosen, provide the basis for selecting the anount of stean flow in excess of that pennitted by the steau flow limiter.
Q211.167 The depressurization rate has a proportional effect on the voiding (15.1. 3. 3.3) action of the core. For the " pressure regulator failure-open" transient, the asstmed depresstrization rate results in a L8 trip.
The results are not consistent.with GESSAR 238-732 where a lower depresstrization rate results in a trip frcm low tu-bine inlet pressure. Explain this discrepancy and provide justification that the asstmed trip grovides the most restrictive margins on MCPR and peak vessel presstre.
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C211.168 For the " inadvertent opening of a safety / relief valve" transient, (15.1.4.
include the time at which suppression pool temperature alanns 2.1.1) and Technical Specification limit are attained in event Table 15 1-5.
Q211.169 Modify Table 15.0-1 as follows:
(15.0) a) Provide a calculated MCPR value for all events in Table 15.0-1 where a MCPR value is not specified.
b) Correct the following discrepancies between values of parameters in Table 15.0-1 and corresponding text values and confinn other discrepancies do not exist.
Maximun Core Average Surface Heat Flux, SRV
% of Initial Actuation Event Table Text Table Text No Yes 15.2.6 (Case 1) 15.4.4 146.6 80.6 15.4.5 141.0 79.0 (Case 1) 15.4.5 134.6 75.0 (Case 2)
Q211.170 For event category 15.3 in Table 15.0-1, identify the most (15.0) limiting anticipated transient for MCPR and maximun vessel pressure.
Q211.171 Provide an analysis of the " loss of instrunent air" transient.
(15.0) 0211.172 In the description cf event sequences for LOCA inside containment, (15.6.5.2.t} several items need additional clarification.
a) The initiating times for MSIV closure and ECCS actuation in the text description appear inconsistent with the corresponding event occurrence times in Table 6.3-1.
Explain these apparent discrepancies.
b) Confirm that the zero reference time for Tables 6.3-1 and 6.2-8 are the sane.
Q211.173 Add the " initial core cooling" safety acticn indicated in NSOA (15.1. 3. 2.1) Figure 15A.6-23 for the " pressure regulator failure-open" transient to event Table 15.1-4 for consistency.
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Q211.174 The treatment of uncertainties associated with SRV setooints (15.2) appears to be handled in three different ways for the events associated with the sections shown below-I Section Treatment of SRV Setooint Uncertainties 15.2 3 3 4 Setpoints include errors (hign) for all valves 15.2.434 Setpoints are assmed 15 psi higher than the i
valves naninal setpoint 15.2.5.3.4 Setpoints are assuned at upper limit of Technical Specifications fcr all valves.
Explain this apparent discrepancy.
If no discrepancy exists, standardize the wording between these secticns for consistency.
1 Q211.175 It is indicated that the " pressure regulator-closed" transient (15. 2.1. 2.3) with failure of the backup pressure regulator is less severe than the " turbine trip wit.h bypass" transient in Section 15.2 3 This agrees with GESSAR 238. As a result, only a qualitative evaluation of the transient was provided. However, quantitative results fran the Grand Gu2 f FSIR indicate the opposite. The staff's concern is that quantitative results for this transient
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may be similar to those for Grand Gulf. Provide a quantitative analysis of the " pressure regulator-closed" transient assuning failure of the backup pressure regulator.
Q211.176 Iri Table 6 3-3, it is indicated that the corewide metal-water (15.6.5 3.3) reaction for WNP-2 has been calculated at 102% of licensed core power. Explain *y the above calculation was not based on the thennal power of 3462 MWt specified in Table 6.3-2 (104.18% of licensed core power) to be consistent with the thermal power value used for LOCA calculations inside contairrnent.
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Q211.177 Review of the " loss of all feedwater flow" transient indicates (15.2.7.2.1)' that the feedwater flow decreases to zero in 5 seconds. For the analyses presented in the FSARs indicated below, the reactor vessel water level decreases to the L3 scrat trip setpoint as follows:
Time at which Vessel ID, L3 trip occurs, in./no. of Rated Power, FSAR see fuel assemblies MWT Susquehanna 4.6 251/764 3293 Fermi-2 6.8 251/764 3293 Grand Gulf 4.1 251/800 3833 WP-2 7.36 251/764 3323 For WP-2 analysis, it would appear that the L3 setpoint would be reached at a time slightly less than that for either Susquehanna or Fenni-2 because the power level is slightly higher and all three have the same size vessel. Provide an explanation as to why the L3 setpoint for WP-2 should not be attained before that for Susquehanna or Fermi-2.
Incitde appropriate design considerations (differences in piping, setpoints, etc) in the response.
Q211.178 A turbine stop valve full-stroke closure time of 0.10 seconds is (15.2.3.3.2) used in the analysis of the " turbine trip" transients.
Demonstrate quantitatively or provide references that show that turbine stop valve full-stroke closure times snaller than 0.10 second do not result in unacceptable increases in ACPR and reactor peak vessel pressure for transients analyzed in Section 15, or provide either (1) justification that a snaller full-stroke closure time cannot occur or (2) a minimtm full-stroke closure time that will be incorporated in the Technical Specifications.
0211.179 The " closure of all MSIVs" transient (closure time 3 sec) results (15.2.4.3.2) in a position switch scram at 0.3 second and indirectly causes a scran trip of the main turbine and generator due to the decrease in p essure sensed by the main turbine. From Figtre 15.2-5, it cannot be detennined whether or not a turbine stop valve and turbine control valve scram occtrs dtring the time interval that the MSIVs are closing from the full open position to the 90%
scram position. Indicate in Table 15.2-5 the time at which the above indirect scram trips occur and the times at which the TSVs and TCVs become fbily closed.
Q211.180 Provide a detailed discussion of activity above the suppression (15.X) pool, activity releases to the environs, and offsite radiological doses for the bounding transient or accident. In addition, provide justification for the selection of the bounding transient.
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Q211.181 For the " loss of AC power" transients, it is indicated that the (15.2.6 3.4) trip of the feedwater turbines may occur earlier than simulated if the inertia of the condensate and booster punps is not sufficient to maintain feedwater punp suction p essure above the low suction pressure trip setpoint. 'Ihe simulation of this transient asstmes sufficient inertia and, thus, the feedwater ptmps are not tripped until the time that level reaches the high water level trip setpoint (L8). What quantitative effect on MCPR and peak vessel pressure would an earlier trip (insufficient inertia) of the feedwater turbines have?
Q211.182 Revise Table 15.2-12 to indicate the time that suppression popl (15.2.9.2.1) alanns are received, the Technical Specification limit is exceeded, and the maximm value of the suppression pool temperature is attained.
Q211.183 In the analysis of the one and two recirculation ptmp trip events
( 15. 3.1. 3.2) in Section 15.3.1, a minimtm design rotating inertia was used to obtain a predicted rate of decrease in core flow greater than expected. Specify the inertia value used for each applicable transient in Section 15 and the basis for selection. Discuss the sensitivity of MCPR and peak reactor vessel pressure to changes in the inertia value.
Q211.184 Frcm the text description in the Grand Gulf FSAR, it is indicated (15 3 2.3 3) that the design of the hydraulic limit on maximtm valve stroking rate it intended to make the fast closure of one and two ree.irculation valve transients less severe than the corresponding trip of one and two recirculation ptmp transients in Section 15.3.1.
However, the results for events 15.3.1 and 15.3 2 in Table 15.0-1 indicate that for the one valve case this does not occur for WNP-2.
a) Explain *y the transient result for the one valve closure event in Section 15.3.2 is more severe than the result in Section 15 3.1.
b) Explain *y a scram occurs for the analysis of the " fast closure of one recirculation valve" transient in the WNP-2 FSAR in view of the fact that for the same analysis presented in the Grand Gulf FSAR, no scram occurs.
Q211.185 For the recirculation ptmp seizure eccident we note in Table (15 3 3) 15 3-5 that credit is taken for nonsafety-grade equipment (L8-trip) to terminate this design basis accident (DBA). Section 15.3.3 of the Standard Review Plan requires use of only safety-grade equipment to mitigate the consequences of this DBA and that the safety functions be acccmplished asstming the worst single failure of an active ccmponent. Re-evaluate this DBA with the above specific criteria and provide the resulting ACPR, peak vessel pressure, and percentage of fbel rods in boiling transition. Asstne coincident loss of offsite power as required by the Standard Review Plan.
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Q211.187 7he narrative on page 15.4-19 for the "startup of an idle (15.4.4 3 3) recirculation punp" transient indicates the core inlet flow rises sharply shortly after the punp starts. However, Figure 15.4-6 does not show this sharp change. Explain.
Q211.188 A maximun stroking rate of 30%/second and 115/second was used for (15.4.5.3.2) the fast closure of one and two recirculation control valves, respectively, in this section and for the events in Section 15 3 2.
In the description of the recirculation control valve stroke rate in Appendix H.3 3.3.7.3.1, the bases for the above stroking rates are not provided. Provide supporting data to l
justify how the above stroking rates for the analysis of events in Sections 15.4.5 and 15.3.2 were obtained.
Q211.189 For the "failtre of RHR shutdown cooling" event, specific input (15.2.9 3) parmeters for the models used to evaluate blowdown rate and suppression pool temperature are shown in Table 15.2-13 along with the analytical results in Figures 15.2-16, -17, -18, and -19.
In connection with this, provide the following infonnation:
a) Identify the analytical models used to evaluate blowdown rate and suppression pool temperature.
bl Revise Table 15.2-13 to include all the input paraneters for the models to be identified in step a) and !rovide justification that the input paraneters are conservative.
In addition, it is indicated that only a qualitative evaluation of the " failure of RHR shutdown cooling" transient is provided because the core behavior has been analyzed in Section 15.2.6.
Update the FSAR to indicate a quantitative analysis has been provided.
Q211.190 Explain the following frcm Figtre 15.4-7, " Fast Opening of Main (15.4.5.
Recirculation Loop Valve at 30% Per Second":
331) a) What causes the drive flow to exceed 100% of rated and level out?
I b) Why doesn't the core inlet flow exceed 100% of rated as a result of the drive flow exceeding 100% of rated?
Q211.191 What causes the core inlet flow and drive flow to exceed 100% of (15.4.5.
rated in Figure 15.4-8, " Fast Opening of Both Recirculation Loop 3 3 2)
Valves at 11% Per Second"?
0211.192 Prgvide justification for use of a HPCS injection temperature of (15.5.1 3 2) 40 F in analysis of the " inadvertent HPCS startup" transient.
Referenced sttdies should be specified.
3 Q211.193 From the discussion of single failtres for the " inadvertent HFCS (15.5.1.2.3) startup" transient, it is indicated that a single failure of the pressure regulator or level control will aggravate the transient, resulting in reduced thermal margins. Provide the MCPR and peak vessel pressure values that result for this event with the most i
limiting of the above single failures considered in the analysis.
Q211.194
'Ihe response to Question 221.02 indicates that 8 x 8 fuel bundles (15.0) with two water rods will be used at WNP-2 instead of the 8 x 8 fuel btaidles with one water rod, a) Have the transients and accidents in Chapter 15 been evaluated with 8 x 8 fuel bundles using one or two water rods?
b) If the transients and accidents in Chapter 15 were analyzed with the one water rod fuel bundles, what changes in MCPR, peak vessel gressure, percent of rods experiencing boiling transition, and the radiological consequences will result if the two water rods design is used in the analyses?
0211.195 In connection with parmeters and asstaptions used for LOCA (15.6.5 3.2) calculations inside containment, provide the following items to aid the staff in evaluating their conservatisn.
a) An explanation as to why a MSIV clostre time of 3 5 seconds in Table 6.2-3 was chosen. Elsewhere-in the FSAR, either 3 seconds or 5 seconds were used in analyses.
b) Explain why the core heatup calculation in Table 6.3-2 asstnes a btaidle power consistent with operation of the highest powered. rod at 102% of the maximta (technical specification) linear heat generation rate (LHGR) instead of operation at 104.18% of the maximtm LHGR which is equivalent to a core thermal power of 3462 MWt.
c) Explsin W1y the core thermal power value of 3462 MWt in Table 6.2-4 is indicated as 102% of licensed core themal power (3323 MWt) instead of 104.185.
d) A tabulation of all permitted axial power shapes addressed by LOCA calculations inside containment. Identify the least favorable axial shape (most conservative) associated with each break size and provide justification of its conservatism.
Q211.196 Operating experience has shown that where thermocouples are used (6 3) to verify ADS valve operation a " false" temperattre increase may be indicated even thotsh the valve has not operated. A direct indication of valve position or flow must be used. Specify how you will meet this requirement.
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C211.197 Section 6.3.2.2.1 of the FSAR states that the HPCS system will (6.3) autanatically switch over fran the condensate storage tank (CST) to the suppression pool if the CST water supply becenes exhausted or is not available.
Review of Figure 7.3-10o indies.tes that automatic switchover will only occur if the CST water level drops to the minista level and activates any one of the four level switches (two per tank). However, in the event that CST water cannot be supplied to the ptap while the CST water level is above the minimun water level, automatic switchover is preclufed.
Resolve this apparent discrepancy between the P& ids and Section 6.3 2.2.1.
Q211.198 Expand the discussion in Section 6 3 to describe the design (6.3) provisions that are incorporated to facilitate maintenance (including draining and flushing) and continous operation of the ECCS ptmps, seals, valves, heat exchangers and piping runs in the long tenn IDCA mode of operation considering that the water being recirculated is potentially very radioactive.
Q211.199 Disetas the design provisions that permit manual override on the (6 3)
ECCS subsystems once they have received an ECC3 initistion signal.
Also, incitde a discussion of any lockout devices or timers that prevent the operator fran prematurely terminating ECCS functions.
If there are plant grocedtres to cover this situation, indicate briefly what instructions are provided.
0211.200 Provide isometric drawings of the major piping for each ECCS (6 3) subsystem (i.e., LPCI, LPCS, etc) to aid in the evaluation of NPSH and possible equipnent flooding. These drawings should show relative elevations and physical locations of the valves, suppression pool, primary contairment, ptaps, heat exchangers, and the lengths of ECCS piping. The location and ntaber of each of the major valves should be shown on the isometric drawing.
0211.201 Several plants have used sandbags or sand-filled tanks as (6 3) biological shielding inside containment.
In the event of a LCCA, these tanks or bags could be damaged and sand could be released.
Release of sand inside contairment cottld result in damage to the ECCS ptaps.
Identify any areas where sandbags or sand-filled tanks are used for biological shielding. Wcat precautions would be taken to prevent ECCS danage if sand or ninilar material were released within containment?
.a Q211.202 A timer is used in each ADS logic. The time delay setting before (6.3) actuation of the ADS is long enotsh that the HPCS system has time to operate, yet not so long that the LPCI and core spray systems are tnable to adequately cool the fuel if the HPCS system fails to start. Manual reset circuits are provided for~the ADS initiation signal and primary containment high pressure signals. By resetting these signals manually, the delay timers are recycled.
The operator can use the reset pushbuttons to delay or prevent automatic opening of the relief valves if such delay or prevention is necessary. The operator may also interrupt the depressurization at any time by the same action. The operator would make this decision based on an assessment of other plant conditions.
Discuss in detail any criteria to be given to the operator (e.g.,
in energency procedures, or operator training) that would fonn the bases for the operator's decision. Discuss the consequences of interrupting ADS depressurization prior to reaching the injection pressure for low pressure systems.
Q211.203 Restricting orifices are conmonly installed downstream of a punp i
(6. 3) to limit the maximun flow rate that could occur and prevent punp danage if the punp discharge line were to fail (i.e., punp runout protection). It is not clear Wiether or not restricting orifice plates will be used for the LPCI system at WNP-2.
Figures 5.4-13a and 5.4-13b show a restricting orifice in the injection piping of each LPCI loop. However, note 9 on Figure 5.4-13a states that these orifices are recommended but not required.
Describe precautionary measures taken to reduce the potential for LPCI punp damage due to rtmout conditions.
i Q211.204 Figures 6 3-53a, -53b, -54a, and -54b show the results of a break (6.3) in a core spray line from the " lead plant" analyses. The assuned single failure shown on the figures does not appear to be the most limiting.
It muld appc;r that the LPCI diesel-generator failtre (division 2) would be more restrictive than the LPCS diesel-generator failtre (division 1), i.e., only LPCI loop A would be available to reflood the core. Explain why failure of the LPCI diesel-generator (division 2) does not result in a higher peak cladding temperature than'that shown en Figure 6.3-54b.
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4 0211.205 Resolve the following discrepancies or inconsistencies:
(6.3) a) Table 1.3-3 and Figure 6.3-2 state that the HFCS system will deliver 6350 gp at a differential pressure (vessel to pmp suction) of 200 psid. Table 6.3-2 indicates that HPCS will deliver 6250 gp at the same differential pressure.
b) Table 13-3 and Figure 6 3-6 state that the LPCS system will deliver 6350 gpn at a differential pressure (vessel to drywell) of 128 psid. Table 6 3-2 indicates that the LPCS system will deliver 6250 gpn at a differential pressure of 122 psid.
c) Table VI of Figtre 5.206 indicates that the top of the active core is 360.3 inches above vessel zero. Figure 5 3-2 indicates that the top of the active core is 366.31 inches above vessel zero.
d) Subsection 6.3.2.2.4 of the FSAR does not mention the relief valves (F088A and F088B) that are installed on the suppression pool suction pipes for loops A and B. -These valves are the same size as the loop C valve (F088C). See the response to Q211.027 and Figures 5.4-13a and 5.4-13b.
Q211.206 The ECCS discharge line fill syr.tems require additional clarifi-(6.3) cation. Provide the jockey ptsop characteristics (head, capacity, etc) and the maximun expected leakage rates for each system discharge piping.
0211. 3 7 Subsection 5.2.2.10 of the FSAR states that the manual and (6.3, automatic actuation of the relief mode for each safety / relief 5.2.2) valve is to be verified in preoperational testing. Subsection 6 3.4.2.2 of the FSAR states that each individual ADS valve is manually actuated prior to or following a refueling outage. The spring setpoint (safety mode) of each valve is to be checked dtring bench tests during refueling outages. On what schedule will safety / relief valves, other than the ADS valves, be manually operated in the relief mode to verify that the valve is operational?
How many of the safety / relief valves will be removed during each refuelin6 outage to receive preventive maintenance and be tested?
C211.208 Appendix A to Regulatory Guide 1.68, Rev 2, sunmarizes the systems (6.3) to be tested and the perfomance capabilities that should be demonstrated by each NR applicant during the preoperational and initial test prcgrams.
It is unclear if the ECCS subsystems are tested using normal and energency power supplies.
Provide assurances that both the nomal and emergency power supplies are used to verify ECCS operability.
If emergency power is not to be used in the operability tests, justify the exception to the criteria of Regulatory Guide 1.68, Rev 2.
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Q211.209 Provide assurance that yotr relief valve design is qualified (5.2.2)
(including testing after being subjected to an enviroment (6 3) representative of an extended time period at normal operating conditions) to support your asstmption that six of the seven ADS valves will operate. A quantitative history of safety / relief valve operation, including similar valves in other plants, should be incitded in this evaluation.
Q211.210 he response to Q211.088 is unacceptable.
It is indicated (15.2.2) that because the generator load rejection transient is not the most limiting transient, the anall increase in surface heat flux that ocetra for TCV closure times of less than 0.15 seconds will not affect the MCPR operating limit. Because reclassification of the generator load rejection transient to a moderate frequency event may result in it being the most limiting transient, even with reanalysis by ODYN, the effect of TCV closure times of less than 0.15 seconds should be reconsidered in the derivation of the MCPR operating limit.
Q211.211 Re respense to Q211.092 is unacceptable. Explain why the (15.3 3)
DBA-LOCA event is indicated as conservatively bounding the punp seizure event when different acceptance criteria are used for each. D e punp seizure event is evaluated based on exceeding 10 CFR 100 guidelines whereas the main criterion for evalyting the DBA-LOCA event is a peak cladding temoerattre of 2200 F.
Coordinate this request with Q211-94.
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g NUCLEAR REGULATORY COMMISSION
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REGION V o,
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SulTE 202, WALNUT CREEK PLAZA
- ,,,e WALNUT CREEK CALIFORNIA 94596 October 17', 1980
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b U.S. NUCLEAR REGULATORY COMMISSION 0FFICE OF INSPECTION AND ENFORCEMENT Headquarters / Region V NOTICE OF SIGNIFICANT LICENSEE MEETING Name ~of Licensee: Washington Public Power Supply System Name of Facility: Washington Nuclear Project Unit 2 Docket No.: 50-397 Date and Time of Meeting: October 21, 1980 10:00 a.m.
Location of Meeting: NRC - Region V Office, Walnut Creek, CA Purpose of Meeting: Licensee provide information regarding restart of construction installation activities at WNP-2 IE Attendees:
R. H. Engelken, Regional Director, RV G. S. Spencer, Chief, RC&ES Branch R. C. Haynes, Chief, Reactor Projects Section R. T. Dodds, Chief, Engineering Support Section D. P. Haist, Reactor Inspector A. D. Toth, Resident Inspector Licensee Attendees:
W. C. Bibb', Project Manager R. M. Foley, Engineering Director B. A. Holmberg, Change Manager R. T. Johnson, Project Quality Assurance Manager A. Sastry, Quality Assurance Manager R. G. Matlock, Managing Director's Office Distribution:
H. D. Thornburg, Director, IE:HQ R. H. Engelken, Director, RV D. Thompson, X00S, IE:HQ G. S. Spencer, RC&ES Branch Chief R. L. Tedesco, Assistant Director, DL, NRR j
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UNITED STATES i
g-NUCLEAR REGULATORY COMMISSION
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SSINS #0660 OCT 2 21980 pc'
!EMORANDUM FOR: Norman C. Moseley, Director Div.ision of Reactor Operations Inspection, IE FROM:
Edward L. Jordan, Assistant Director for Technical Programs, Division of Reactor Operations Inspection, IE
SUBJECT:
FORTHCOMING MEETING WITH BURNS AND RGE REGARDING SEISMIC QUALIFICATION OF ELECTRICAL EQUIPMENT Date and Time:
Friday, October 24, 1980 9 :00 a.m.
Location :
Phillips Building P-110
Purpose:
To discuss the seismic qua'lification of electrical cabinets with rigid conduit attached and tests performed by Wyle Labs for WPPSS No. 2.
Participants :
NRC Burns and Roe T. Westerman T. Hendrickson H. Wong C. Hofmayer P. Y. Chang Edwa L. Jordan, Assistant Director fo Technical Programs Division of Reactor Operations Inspection CONTACT:
H. J, Wong, IE 49-28180 50n eg o n no _
3 OCT 2 21980 e.w W. -ryp1 MEETING NOTICE DISTRIBUTION bo 930, otfI 9,',9. s: \\, -
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NRC PDR '
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F HDenton ECase DEisenhut RVollmer Dross SHanauer FSchroeder TMurley GArlotto TNovak DLynch RTedesco Glainas JKnight VNoonan ZRosztoczy PCheck LRubens tein MErnst LShao GMadsen, RIV UPotapous, RIV DFox, RIV BDodds, RV HThornburg WRutherford ELD Receptionist - Phillips Bldg.
NRC Participants TERA ACRS (16) e
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