ML20198G070
| ML20198G070 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 10/18/1974 |
| From: | Stello V US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Moore V US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| CON-WNP-1006 NUDOCS 8605290224 | |
| Download: ML20198G070 (10) | |
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't Docket' Files'(
L Reading File
- ~~C RSB Reading File OCT 16 131.;
Docket Nos, 50-460 and 50-513 1
Voss A, Moore, Assistant Director for Light Water Reactors Group 2, L PSAR SECCND ROUND REVIEW Plant Name:
WPPSS 1 & 4 Licensing Stage:
CP Docket Nos,:
50-460 & 50-513 Responsible Branch LWR 2-3 and Project Manager:
T, Cox Technical Review Branch Involved:
Reactor Systems Branch Requested Completion Date:
October 10, 1974 Description of Review:
Second Round Questions Review Status:
Second Round Complete The Second Round Review of the Wisconsin Public Power Supply PSAR was completed by the Reactor Systems Branch, This evaluation reflects our review of the application through Amendment No, 10.
The enclosed list summarizes our positions on the information available, In addition, the PSAR must be supplemented to address ECCS (Appendix K) and ATWS considerations, The 17x17 Mark C fuel bundle design is being reviewed generically, When completed, the results of this review must be incorporated into the PSAR, Ort!nal :lin6J by V. Stel!a Victor Stello, Jr., Assistant Director for Reactor Safety Directorate of Licensing
Enclosure:
Staff Position ec:
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212,0 SECTION A - REACTOR SYSTEMS BRANCH
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212,1 The applicant should commit to a schedule for the completion (1,5) of all R&D programs for the 17x17 fuel so that a timely g
review of this information can be performed during the e
operating license review period. A description of the
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additional required tests and an implementation schedule for the entire R&D program should be included or referenced in Section 1,5 of the SAR, s
l 212,2 The reference to' Figures 4,2-3 and 4,2-5 should be corrected I
(4,2,2,2) or the figures should be updated to support the -
j discussion in this section, 212,3 The radial-local peaking factor for the most probable i
(4,4,3,4) design is listed as 1,54 on page 4,4-27.
It'is the j
Staff's understanding that the correct v,tlue should i
be 1,53 as specified on page 4,4-2, Correct this inconsistency, 212,4 It is the Staff's position that Section 4,4,4 or (4,4,4) 4,4,2,8,2 should be amended to include a commitment to de..snstrate that the BAW-2 critical hsat flux correlation (CHF) will conservatively predict the 17x17 CHF data, We require that the DNB correlation used for design results in a 95% probability, with 95% confidence, that critical heat flux will not occur.
212,5
'It is our position that Section 5,5,13 should be expanded (5,5,13) to include the relieving capacities, set points, and set-point tolerances of the pressurizer safety and relief valves and all other pressure relieving devices in all systems connected to the reactor coolant' system, Provide the bases on which these capacities were chosen, Describe how the. flow rates of these valves are verified experi-mentally, Include manufacturer's certification testing and applicant 's preoperational and operational testing that will be required to ensure that the valves will operate properly at the specified conditions.
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~6, 212.6 The Staff takes-the position that analyses should be provided (6,3,2)
'to show that cavitating venturis are not recuired in the HPIS piping, " Assume that the break occurs in the high pressure ECCS piping between the cold leg and the check valve and that the diesel that powers the HP pump in the other train fails to operate, This resuits in partial flow from the pump in the broken train reaching the core, Demonstrate that this flow l
is adequate to mitigat e the consequences of the postulated small break in the high pressure piping.
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l 212 7 Refer to Section 6.3 4, Tests and Inspections, This section j
(6 3,4) shall be amended to demonstrate that all emergency core cooling system component and related systems tests are per-a formed in accordance with Regulatory Guide 1,79 All i
exceptions must be clearly identified and explained, l
l 212,8
It is the Staff's. posit' ion that design provisions should be (634) made to fac'111 tate maintenance and continuous operation of the ECCS pumps,' seals, valves, heat'exenangers, and piping runs in the long-term cooling mode bf operation following a LOCA, Discuss these design provisione assuming that the coolant is radioactive.,'
E 212',9 The PSAR should be amended t6, include a complete characterization (6,3 Ey of the core barrel vent valve design. If the Oconee design is referenced, are all the tests and analyses that have been completed for this design equally applicable to the WPPSS design? Include in the core barrel vent" valve description a drawing showing the relative location of the core ficoding nozzles with respect to the vent valves,
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212.10 Refer to request for information 6,62, Changes to the Staff j
(6,3)
. osition concerning the treatment of the core barrel vent valves p
in the thermal-hydraulic analysis will be considered as experience from operating' plants which contain vent valves become available for Staff review. This data should confirm that the valves remain closed during normal operation. If this data are
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available, provide the reference. If not, the Staff position shall remain as stated in request for information 6.62.
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212-3 212.11 For ECCS piping which include check valves for isolation from (6 3 2) the reactor-coolant system, it is the staff's position that leak detection capability be provided to assure that adequate information is available to the operator to effect isolation of a system and prevent overpressurization. Discuss the leak detection capability in the WPPSS design.
212,12 It is the staff's position that adequate information should (6 3 2.14) exist in the PSAR to confirm pressure losses used in the ECCS thermal-hydraulic analyses and NPSH calculations. Provide dimensioned isometric drawing (s) that show the relative elevations and the physical locations of the BWST, the sump used for the recirculation phase of ECCS operation, the core flooding tanks, the HPI and LPI pumps, the DHR heat exchangers, j
and the lengths of piping for the entire ECCS, The location of all valves should be shown on the isometric drawings. The valve nomenclature should be identical to that used on the piping and instrument drawings presented in the PSAR. Also describe the design provisions that preclude freezing of the BWST and its contents.
212,13 The staff takes the position that the ECCS piping between the (6,3 2.10) sumps and pumps should be vented to preclude pump cavitation and reduce load impact to the internals. Discuss the method of venting of the ECCS piping between the sump and the pumps.
Identify any operator actions required to vent this piping prior to initiating the recirculation phase of ECCS operation following a postulated LOCA.
212.14 Expand Subsection 6.3 2.14 to include or reference a detailed (6 3 2.14) description of the experimental procram which will be used to check the adequacy of the pump NPSH for ECC operation at the site.
212.15 In the response to AEC request 6,56:
(6 3.2.14)
(1) it was stated that the DHR/LPI pump performance char-acteristics would be furnished to the staff by July 1974.
This commitment. has not been fulfilled; t
f (2) when calculating the minimum NPSH available for the DHR pumps, what is assumed for the height of the column of water measured from the bottom of the sump? What is the volume a
of water in the cump as a function of height above the bottom of the sump? Demonstrate that sufficient NPSH would be i
readily available to the DHR pumps during the ECCS
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recirculation mode of operation for design and runout i
flow rates.
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212.16 The information contained in the response to request 6.58 is (6.3.2.17)
_not adequate for the Staff to determine (a) if the operator' has ample time from the BWST. low-level alarm to comple te the transfer from injection to recirculation mode of emergency core cooling before the water inventory in the BWST is depleted and (b) the consequences resulting from temporary delay in achieving the switchover.
Provide the following additional information:
(1) It is the Staff's position that the reply to question 6.58 is inadequate and requires a more detailed answer.
The applicant should specify the minimum tine the operator has to switch from the injection to recirculation mode of
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ECCS operation af ter the annunciation of the low-level BWST alarm for each of the following conditions:
(a)
Large pipe break with the HPI, LPI, and BS pumps operating at maximum capacity, (b) Small pipe break where high pressure recirculation through the LPI pumps is required.
Discussion should include control board layout, alarm redundancy, valve interlocks and operating procedures to justify the adequacy of this time interval.
(2)
Provide an analysis showing the core coolant mixture level, system pressure, and maximum fuel temperature as functions of possible delay times for the hypothetical situation where the recirculation mode of emergency core cooling is delayed.
The maximum celay time considered in the analysis can be related to the temperature limits as specified in the " Acceptance Criteria," dated January 4, 1974, in the Federal Register.
(3)
Discuss the potential for pump damage and other possible consequences resulting from inadvertently switching to the sump when inadequate vater supply exists in the sump.
Discuss the interlocks and other system designs to preclude operation in the recirculation mode without adequate water inventory in the sump.
(4)
Assuming that no operator action takes place or takes place after the LP/DHR pumps cavitate, discuss your reasons for not providing an automatic system to complete the switchover or some other protective means-to preclude damage to these pumps. We note that other B&W plants (Davis Besse 1, Bellefonte, Greenwood, and B-SAR-241) have included an automatic switchover system to increase reliability and maintain ECCS integrity.
212.17 After careful review of your response to AEC request number 6.51, (6.3.2.17) the Staff have not modified their position concerning the valves (V63A and V64B) in the two lines between the HP pump suction headers and the DH pump discharge lines. These two valves should
'l be remote manual valves with indication and control from the control room to facilitate system alignment as required for intermediate or small breaks.
This requirement has been incorporated I
into the Greenwood and Bellefonte designs.
3 212-5 212,18 It is our position that analyses be providec to show the (6,3,2,18) bases (including a description of the transients considered)
/y on which the relieving capacities of the valves shown in Table 6 3-6 were chosen, Present the results of analyses
.1 of the most limiting transient to'show that relief valves VB3-A and V84-B protect against overpressurization of the LP/DHR system Describe how the flow rates of these valves are verified experimentally, Include manufacturer's certification testing and the suggested applicant's pre-operational and oper.ational testing that will be required 6
to ensure that the valves will operate properly at the specified conditions, 212,19 The response to request 6,63 for additional information on the (6,3,2,19) cavitating venturis is incomplete. Provide the following additional information:
(1) Test data were provided to demonstrate that the venturi will deliver a constant flow rate for various downstream i
pressures, Provide more detail on the test procedure, number and designs of venturis tested, how the test modelled the LPI system, pump characteristics, and instrumentation, I
(2) If the test conditions and venturi models were different from PWR conditions where the venturis are expeuted to function, describe in detail the technique used to scale the, tests.
(3) Discuss how the venturis in the LPI system will be tested in the reactor during preoperational testing to ascertain that flow splitting will occur. How will the break and system conditions be simulated for the LOCA?
(4) For the design flow specified in Table 6.3-1 for the LP/DHR pump, were the resistance and pressure drops for the venturis included? Do the venturis restrict the, flow such that a larger pump head is required to achieve design flow when compared to the same system without the venturis?
(5) It is our position that the PSAR should be amended to include a section describing the venturis in the LPI system.
Drawings, analyses, results of tests,for calibration and flow division, bases for instrumentation, and other information pertinent to the speelfication of the venturis should be included.
(6) Considering the inlet temperature and pressure of the water from the BWST, it is not clear that flashing will occur un-less the throat area is extremely small, Cavitation also appeart to be functions of the pump characteristics and the downstream fluid properties. Provide detailed analyses using anticipated LPI system conditions, venturis designs, I
and system piping pressure drop as a function of flow rate to demonstrate that flashing will eccur,
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212.20 Refer to Figure 3.2-4, Decay Heat Removal System P & ID.
(6.3.3)
Each pump suction line from the hot Icg contains.two motor-operated valvca inside containment and a manual locked-closed valve outside containment, Where is this yalye located?
If the valves in both Ictdown lines are in the same proximity, it is the Staff's position that no conditions (e.g., high radiation levels) shall exist that could prevent manually opening the valves by the operator. Confirm that the operator will always be able to open these valves.
212.21 It is the staff's position that the water inventory in the BWST (6.3.3) be sufficient to satisfy emergency core cooling requirements i
during the recirculation mode of ECC'S operation to provide long term decay heat removal.
Is all of the water in the BWST ultimately available for recirculation from the sump f
following a LOCA considering both ECCS and core spray operation?
If not, provide a list of the quantitics of water that are un-available and their ultimate locations within the containment.
Your response should consider any water that ultimately fills the reactor vessel cavity and refueling canal.
212.22 The AEC " Interim Acceptance Criteria" has been superseded as stated in the Federal Register, Vol. 39, No. 3, Friday, January 4,1974.
It is required that analysis.and evaluorion of ECCS, cooling performance following postulated loss-of-coolant accidents shall be performed in accordance with the requirements of 10 CFR 50.46 using an Evaluation Model in conformance with Appendix K.
A commitment is required from the applicant identifying when the Safety Analysis Report will be revised and re-submitted so that the review may proceed. This commitment should appear in Subsection 6.3.3.1.
212.23 The staff position is that for each accident and transient (15.0) analyzed in Chapter 15 of the SAR which results in a reactor trip, the control rod worth curve (reactivity vs. time) used in the analysis must be provided. Describe the axial flux shape, used to generate this curve.
In addition, provide graphs showing control rod positions as a function of rime and core reactivity as a function of control rod position.
Explain why the axial flux shape and peaking factor used to generata the control rod worth curve is considered conservative. Alternatively, the control rod worth curve can be shown to be consistent with the axial power profile assumed for the accident or transient so that jointly they. produce the worst consequence. Relate the axial power profile and control rod worth curves to the opera-tional imbalance limits and control rod limits specified in the Technical Specifications. Reference the Technical Specifi-cations which require the operator to maintain this flux shape over the life of the core and the maximum allowed variation permitted in this flux shape.
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- i"j 212.24 It is the staff's position that the bases be provided for 3
(15.1) the assumption that the uncertainties in the core inlet j
temperature and initial system pressure will not exceed 3*F 4
and -45 psi, respectively.
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I 212.25 For the various combinations of reactor coolant pump operating i
(15.1) conditions proposed in Section 4.4, it is the staff's position
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that the worst-case initial conditions with respect to the number of operating reactor coolant pumps be established for each of the events in Chapter 15.
212.26 It is the staff's position that a list be provided of all i
(15.1) the transients and accidents considered in establishing the auxiliary feedwater system configuration and capacity. For i
cach of these transients and accidents, state the maximum j
delay in the initiation of auxiliary feedwater flow that can be tolerated, whether the initiation is automatic or requires operator action, and the minimum required auxiliary feedwater flow rate required to mitigate the consequences of the transient or accident.
212.27 For the engineered safety systems, the staff position is that the applicant must provide lists of the equipment assumed to function after each accident or incident evaluated in Chapter 15.
Include the ESF equipment required for long-term heat removal and the switching from ECC injection to ECC recirculation under all anticipated conditions.
Indicate for each component whether power is required and available from norcal AC sources, emergency AC power, or DC batteries.
j 212.28 It is the staff's position that Subsection 15.1.2, uncontrolled 1
control rod withdrawal at power, be amended to include DNBR results to show that the corresponding limits in Section 4.2 were not exceeded. These histories should be presented as a function of the reactivity insertion rates considered.
Include the pressurizer 1cvel and spray rate as a function of time.
If the pressurizer goes solid, provide a discussion of the method used for calculating water relief rate through the safety valves.
212.29 In response to Questions 15.11, it was stated that the maximum (15.1.4) dilution flow rate of 200 gpm was based on the maximum letdown flow rate.
If it is possibic for the makeup flow rate to exceed this because of operator error or a single failure, describe the resulting accident at the maximum makeup flow rate with the reactor in both manual and automatic control. Relate the severity of this accident to the results previously given as to maximum overpower, pressure increase, and time to zero shut-down margin. List the assumptions used to define the maximum makeup flow rate.
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212-8 It is the staff position that the shutdown margin availabic at the time of the accident should be the minimum allowed by the Technical Specifications.
State what this minimum j
is and how long it takes to reduce it to zero at the maximum 4
dilution rate.
I 212.30 It is the staff's position that additional information is (15.1.8) required to properly assess the results of the feedwater e
line rupture accidents. Provide the following:
1.
Analyses for a feedwater line rupture inside and outside containment. In the analyses justify the method used to calculate break flow, the sizes and locations of breaks.
Show that the single failures considered in the analyses are the most limiting ones. Further, if the pressurizer goes solid as a result of the accident, provide bases for water discharge rates through safety j
valves.
2.
As a function of time, plots of steam generator unter level, DNBR, safety valve flow, auxiliary feedwater flow, steam generator pressure, primary pressure, steam flow, and steam generator masses.
3.
List in chronological order all operator ' actions.
dre the steam generators isolated using the steam line 4.
isolation valves or is credit taken for the closure of the turbine stop valves? Justify your assumption.
212.31 (15.1.17)
Based on the response to request 15.18, concerning the steam-generator tube rupture accident, it is not clear that your hand-calculation method of analyzing this event properly includes all system variables.
formation as a function of time: Provide the following in-DNBR, neutron power, the average and hot spot heat fluxes, coolant conditions, pressurizer level, steam generators level, feedwater flow rate, and safety relief valves flow races.
For a tube leak rate which is approximately equal to the make-up rate, how would the operator properly diagnose the problem?
Identify the normally operating plant instrumentation and controls which are assumed operable in mitigating the con-sequences of this accident.
4 212.32 With regard to the steamline break in Subsection 15.1.15, it is (15.1.14) the staff position that this subsection be amended to include the following:
1.
Incorporate into the analyses of a steamline break inside
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of containment the consequences of the failure of one main
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steamline isolation valve to close concurrent with a pipe
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212-9 break in one of the other three main steamlines. Jus-tify.the use of any non-seismic Category I components I
(i.e., turbine stop valves) that are used to mitigate j
the consequences of a steamline break.
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j A detailed discussion of all required operator actions and corresponding times of action.
Discuss the potential for and consequences of an operator error allowing auxiliary feedwater to be admitted to the affected steam generator.
1 3.
j Compare the consequences of a double-ended 38-inch main steamline break both inside and outside the containment j
building.
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Provide plots of the following parameters versus time for all break analyses: steamline break blowdown' rates, water j
level in each steam generator, DNBR, emergency core cooling
's flow, pressurizer water level, hot and average channel heat j
flux, and safety and relief valve flow rates.
212.33 It is the staff's position that Chapter 15 include an analysis (15.1.36) of the events following an inadvertent opening of a pressurizer safety valve as required in the October 1972 Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants.
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