ML19289G322
| ML19289G322 | |
| Person / Time | |
|---|---|
| Site: | Crane, Midland  |
| Issue date: | 12/27/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19289G307 | List: |
| References | |
| NUDOCS 7908220503 | |
| Download: ML19289G322 (6) | |
Text
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RSB Ro"nd Two Questions / Positions Midland Units 1 & 2 211.176 The response to question 211.147 provides the initial conditions 5
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f r BAW-10043 to show that it " brackets" the Midland units.
It is not clear that BAW-10043 bounds the Midland units. Comparison (g3p) of parameters from the Midland FSAR and your response is as follows:
BAW-10043 Midland FSAR Core power 3105 MWT (112%)
2452 (100%)
Pump Heat 16 MWT 16 MWT RCS Flow Rate (lb/hr) 137.9x10 126.3x10 6
6 Pressurizer Code Safety 690,000 595,690 Valve Capacity (lb/hr)
Secondary Safety Valve 13,680,000 12,484,520 Capacity (lb/hr)
The effects of less flow and relief valve capacity ar[not obvious relative to the loser power levdl. Submit a plant-specific over-pressure valve sizing calculation for Midland. Also, the analysis should assume that the reactor scram is initiated by the second safety grade signal from the reactor protection system.
211.177 The response to question 211.131 does not satisfy our concarn with respect to the detection and isolation of passive ECCS failures
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(RSp) during the long-term cooling phase after a LOCA. Although the test report addressing injection pump seals indicates that seal integrity was maintained for the conditions under which they were tested, we do not concur with your proposal for LPI seal leakage of 500 ml/ min to serve as the bounding leak rata for a passive failure following a LOCA (valve stem packing or pump seal failure). Operating data indicates that leak rates in excess of your proposal have occurred.
" Sounding" leak rate assumptions on the order of 30-50 gpm have been acceoted by the staff in the past; therefore, we recuire -ha you show that the 7908220503 30 lib
. (6.3)
ECCS equipment layout, room water level detectors and airborne jh7) radiation monitors in the Midland plant meet the criteria listed in question 211.47 assuming leakage rates of this magnitude, or revise your design accordingly.
211.178 The response to our position in question 211.129 does not provide
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assurance that the single vent on the BWST is adequate since:
1.
It is on top of the tank and would be susceptible to blockage due to snow build-up.
2.
No heat tracing is provided on the vent.
3.
Your response does not describe the " screen inside the SWST" which is heated.
We require that a BWST vent configuration be provided which will preclude vent blockage due to icing or snow accumulation.
211.179 The response to question 211.125 states that flow indication j
in the " dump-to-sump" lines is not necessary. Our position is that the operator must be provided with flow indication to confirm that at least the minimum required dilution flow exists subsequent to a LOCA.
211.180 The resconse to question 211.105 states that the alarm provided 5) in the control room to detect a reactor building sump level increase corresponding to 1 gpm leak within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> will be generated by the plant computer. Since the plant computer may not be available at all times during plant operation, we recuire tr.at an alarm be orovided in the control rocm which will be available et all times.
211.181 The resconse to question 211.113 states that extended operation (5*3) of the :ecay Heat pemoval pumps at flows less than 800 gom would result in damage to the pumps.
(This was your basis for not using 2 0 0 \\ \\ l' 6
continued recirculation through the DHR heat exchanger and recirculation line to protect the DHR pumps from closure of a suction valve). Confirm, with basis, that the low pressure injection system will perfom its function in the piggyback mode, since the LPI (DHR) pump flow will be less than 800 gpm.
211.182 The response to question 211.103 does not meet our requirements f6 with respect to check valve leak testing. The proposal to test two valves in each of the Core Flood and Low Pr. essure Injection Lines is acceptable for these systems, however, we require that at least two check valves in each of the high pressure injection lines be tested also. This should be done by classifying these valves as AC in accordance with Section XI of the ASME Code.
211.183 The response to question 211.152 does not satisfy our concern
.6) that dilution events couTd occur at rates less than the makeup flow rate alarm setpoint, and would not be detected. Although these events would take longer than 30 minutes to reach criticality, no indication would be provided from the high makeup flow alarm to alert the operator so he could terminate the event. We require that the operator have adequate time after indication of the event in accordance with the folicwing criteria.
Plant Condition Time Prior to Criticality M er Ina cation Refueling 30 minutes Startup, cold shutdown, hot 15 minutes standby, and power operation Provide assurance that these criteria are met or revise your design accordingly.
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211.184 During the recent review of the loss-of-offsite-power O.2) preoperative test procedure for another B&W plant, a concern was developed regarding the control of OTSG level by the auxiliary feedwater system during the event. Specifically, overcooling of the primary system could result from feeding the OTSG with the cold auxiliary feedwater. The cooldown could be large enough to empty the pressurizer and cause a steam bubble to form in the hot leg high points, which could impede natural circulation and core cooling. Address this concern for the Midland units. Provide the results of an analysis of a loss-of-offsite power assuming the worst-case initial conditions (low power appears to be worst since prograr:med steam generator level is lowest).
Include plots of ste=.m generator level, reactor coolant system temperature and pressurizer level. Discuss your assumptions regarding auxiliary feedwater control. Show that MDNBR will remain above 1.30 and core cooling will not be impaired.
211.185 The response to question 211.157 regarding worst case single failure f r a main steam line break is insufficient. The analysis should 5)
(RSp) consider the following:
1.
Inadvertent atmospheric dump valve ocening 2.
Steam flow through all unisolated lines down stream of the MSIV's (Unit 2). Table 10.3-5 indicates that all lines are not isolated after a steam line break assuming the single failure of 1 MSIV.
3.
Process steam cross-connect valves opening, (see question 211.160) unless pcwer will be removed.
Provide your basis for stating that 1 HPI cumo is the worst single failure with respect to overcooling.
Provide the wors: single failure witn basis for ine acest DNB main steam line break.
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- 211.186 Confirm that the bounding Midland Chapter 1.5 accident and transient analyses consider all events which could occur in
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Modes 1, 2, O and 4 as defined in FSAR section 7. 7.1. 6. 2. 2.
We require that all modes of operation be considered in your analysis (.i.e., Unit 1 NSSS supplying Unit 2 turbine) and specifically defined in the Midland Technical Specifications.
211.187 We met with you on December 11, 1978 to hear your appeal of our
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position regarding the capability of the Midland units to achieve a cold shutdown condition. OQr requirements are listed in question 211.35. Your position was that hot shutdown was a safe shutdown condition and that the Midland units could not reach cold shutdown with equipment previously qualified as safety grade, assuming loss of offsite power and a single failure. We informed you that your position was not acceptable and that more work was necessary to satisfy our requirements. The following additional guidance was provided at our meeting:
1.
Repair and maintenance of equipment is acceptable if reasonably justified.
2.
A easonaole time period to reach cold shutdown would be acceptable.
3.
Additional analysis with respect to seismic qualification of systems may be acceptable.
4 A natural circulation cooldown test is required.
5.
Operating procedures are required.
'de reouire that you provide information to show that the Midland units meet the cold shutdown requirements of cuestion 211.35 as nodified witn this aceitienal guidance.
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Enclesure 2 Responses Not Complete:
1.
Missiles Inside Containment (211.95) 2.
Reactor Coolant Pump Mounting (211.100) 3.
Pressurizer Safety Valves (211.104) 4.
Appendix G Overpressure Protection (211.105) 5.
Pressurizer Heaters (211.112) 6.
HPI Line Break Analysis (211.117) 7.
LOCA - Reactor Coolant Flow Input (211.119) 8.
BPRA and ORA Modifications (211.138) 9.
Decrease in Feedwater Temperature Analysis (211.143)
- 10. Locked Rotor Accident Analysis (211.149)
- 11. Loss of Flow Transient Analysis (1 Reactor Coolant Pump) (211.150)
- 12. Feedwater Line Break (211.153)
- 13. Main Steam Line Break Operator Actions (211.161)
- 14. Main Steam Line Isolation Valve Closure Analysis (211.164)
- 15. Main Steam Li.ne Break.(stuck rod assumption) (211.166-211.175)
- 16. Overpressure Protection Report - SAW-10043 (211.la7) 2043 121 6
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NUCLEAR REGULATORY COMM!sslON I I.k M N O
WASHINGTON, D. C. 20555 JAN 2 91979 Docket Nos.: 50-329/330 MEMORANDUM FOR:
S. A. Varga. Chief, LWR-4, DpM THROUGH:
T. M. Novak, Chief, Reactor Systems Branch, DSS FRCM:
G. M. Mazetis, Section Leader, Reactor Systems Branch, DSS
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION CONCERNING MIDLAND 1 & 2 CONTAINMENT SUMP RECIRCULATION TEST Reactor Systems Branch has reviewed the applicants proposal for the model containment sump recirculation test. To confirm that the proposed test will demonstrate the acceptability of the Midland sump design with respect to vortexing and intake losses, we request that the enclosed questions be forwarded to the applicant.
O Gerald R. !az-.is, Section Leader
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Reactor Systems Branch Division of Systems Safety
Enclosure:
As Stated cc:
R. Tedesco T. Novak S. Israel G. Mazetis S. Newberry
Contact:
Scott Newberry, NRR 49-27341 2043 122 OOPG
~M 6 a g op-3 3 Womeo
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MIDLAND 1 & 2 211.0 REACTOR SYSTEMS BRANCH 211.1 We note that your proposal does not model the :entainment structures outside the trash racks and the resultant far field effects. You imply that the trash racks will suppress any vortex generated in the far field and that this effectiveness has been documented for,
other plants. We are concerned that certain vortices could be fomed in the far field particular to the Midland configuration which could penetrate the trash rack. Provide data / justification to show the following:
1.
Far field vortex formation is unlikely for the Midland containment configuration considering the low velocities that would exist.
2.
The trash racks would eliminate all vortices produced in the far field which could approach the trash rack at various angles.
211.2 Provide justification for proposing all tests at the minimum water level since more severe vortices have been known to form at other levels.
211.3 Discuss more thoroughly the instrumentation and method used to measure and calculate the sump and intake structure pressure losses. Address the accuracy and calibration of these instruments.
211.4 Provide your proposal for the in-plant test which will be used to establish as-built piping losses in the Midland ECCS. Discuss how these tests will be used to confirm that the FSAR NPSH calculation is conservative considering the difference between the test conditions (temperature, ficws, flow paths, etc.) and the worst case pumping modes following a LOCA.
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_2 211.5 Provide the details of your FSAR NPSH calculations for the high and low pressure injection pumps including the head loss calculated for each section of pipe and.the associated L/D, K factors, velocities, Reynolds numbers, etc.
211.6 Provide an additional test or test data to justify testing for vortex formation at prototype Reynolds number.
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