ML20198F322
| ML20198F322 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 01/17/1974 |
| From: | Stello V US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Moore V US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| CON-WNP-0943, CON-WNP-943 NUDOCS 8605280475 | |
| Download: ML20198F322 (9) | |
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Docket File RSB Reading File JAN 171'!74 V. Stello L: Reading File L: Administrative Assistant Docket No. 50-!.60 Voss A. Moore. Assistant Director for Light !!ater Rosetors. Group 2, L PSAR FIRST ROUND RWICJ Plant Name: Vashincton Pubiic Power Supply System,I? nit 1 Licensin;; Stage: CP Docket :;o. 30 M0 Responsible Branch and Project Manager: LiiR 2-3. T. Cox 2equested Corapletion Date: January 11, 1974 Technical Review Eranch: Reactor Systems Description of Boview: First Rour.1 Questions Review Status: Additional Information Roquested A review of the Washington Public Power Supply System. thiit 1 was carried cut to deternine acceptability of the information provided in the PSAR.
The MSSS is a Babcock and Wilcox 3600 ICt pressurized water reactor with 17x17 Mark C fuel. The Standard Format and Content of Safety Analysis hports for Nuclear Power Plants (Revision 1)," dated October 1972 was used as the cuideline for the review.
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- Victor Stallo, Jr.. Assistant Director for Reactor Safety Directorate of Licensing Enclosure Questions cc w/o encl.
A. Girenbusso, L W. C. Mcdonald, L ce w/ encl.
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i 4.0 REACTOR I
4.2 In Section 4.4.2.2 you calculate the maximum average cladding temperature assuming maximum design and most probable design conditions.
For the maximum design condition you use a radial-local nuclear peaking factor of 1.55 and an axial-nuclear peaking factor of 1.67.
For the most probable condition you use cor-responding peaking factors of 1.53 and 1.67.
Explain the dif-forence between the two radial values and the value of 1.54 quoted on page 4.4-27.
Also explain the discrepancy between the axial-nuclear peaking factor of 1.67 as used here and the value of 1.70 listed in Table 4.4-8.
4.3 Equation 4 on page 4.4-9, used to calculate the void fractio,n in the slightly subcooled boiling region, is not the same as used in the Greenwood and Bellefonte PSAR's.
Correct any ty-pographical errors or justify the change.
4.4 Clarify the relationship between q" and qy as defined on page 4.4-11.
4.5 Correct the error in the exponent of the equation defining FS on page 4.4-15, 4.'6 Four operational transients are summarized on page 4.4-21.
With regard to this su= mary and Figures 4.4-12 and 4.4-13:
The min'imum hot channel DNB at 90 percent power is shown a.
to have three valves: viz., 3.07, 2.54, and 2.0.
Explain and correct these inconsistencies.
b.
The minimum hot channel DNB ratio at 100 percent power is shown to have two values; vis., 2.54 and 2.0 (as compared to values of 1,82 in Table 4.4-1 and a minimum of 2.11 in Tabic 4.4-2).
Explain and correct these inconsistencies.
4.7 What calculational models or experimental data are used to derive the design nuclear radial peaks shown in Figure 4.4-8?
!!ow is this Figure related to flux distributions given in Section 4.37 4.8 In Section 4.4.2.10.3.f on page 4.4-23 it is stated that one of two sets of data by Christensen predict an EOL melting temper-ature for UO2 of 3000'F. Substantiate or correct.
4.9 Explcin why Tabic 4.4-1 lists the same hot spot max / avg heat flux ratio (Fq nuc and each) for WNP-1 and Bellefonte when the local heat flux factors (Fq") are different? Ifthe@qnueandmech) numbers arc in fact different, correct the max and Avg heat flux values given on the preceding page.
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4.10 In Section 4.4.3.7 the fuel rod bowing due to transients is esti-mated using a cladding temperature distribution found for a flow blockage.
Explain why such a temperature distribution was con-sidered appropriate for determining the worst bowing to be ex-pected during a transient. We estimate a 10'; reduction in local heat transfer due to the 0.049 inch bowing reported on page 4.4-32.
Was this accounted for in determining the maximum clad temperature of 850*F? What bowing and local reduction in heat transfer was assumed for the locked rotor analysis of Section 15.1.57 In the analysis used to derive Figure 4.4-27, how was clad axial conduction treated, and was DNB propagation considered?
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5.0 REACTOR COOLANT AND CONNECTED SYSTEMS 5.8 Referring to Figure 5.1-1, explain why the pressures listed for locations 1 and 2 are the same.
5.9 In Section 5.2.2 B&W topical report BAW-10043 is referenced con-
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cerning overpressure protection of the WNP-1 reactor coolant system.
For both the Bellefonte and Greenwood reactors a new, yet to be published report concerning overpressure protection is referenced.
Explain why the older report is referenced for WNP-1. Also give the basis for choosing the pressurizer relief valve capacity of 500,000 lb/hr per valve quoted on page 5.2-8.
BAW-10043 does not state valve capacity requirements as a function of rated reactor power level; explain this relationship and describe how it was used in the design of WNP-1, 5.10 Specify the NPSH requirements for the reactor coolant pumps and relate to the requirement of Section 5.3.2 that the pump suction pressure be 250 psig before the pumps are started.
This value apparently differs from Greenwood and Bellefonte. Why?
5.11 The reactor coolant pump drawing (Fiqure 5.5-1) is insufficient for understanding the internal functioning of the pump, the lo-cation and arrangement of the seal mechanisms. etc.
Prnvida e arawing sutt1cient to illustrate all of the pump internal mcch-anisms. Name the pump manufacturer; compare the pump to those used on the other B&W reactors; list the forward and reverse K-factors for this pump and for other pumps typical of B&W re-actors.
5.12 On pages 5.2.18, 5.2.19, and 5.5.2, the potential for pump over-speed during a LCCA is discussed.
Please specify and describe what experimental and analytical programs will be undertaken to resolve this problem and on what basis corrective measures are to be taken.
5.13 Table 5.5-1, referred to in Section 5.5.7.2, does not list major component design data for the DHR system.
Supply the correct table.
5.14 In Section 5.5.7.2, it is stated that the decay heat removal pumps are designed for continuous operation for the period required for refueling.
In Section 6.3, it is stated that these pumps are required for long-term core cooling following a LOCA.
Please specify for what time period these pumps are designed to continuously operate throughout this period.
Re-late this reliability information.(e.g., mean time between failures) to design provisions to facilitate pump maintenance during refueling or post LOCA.
6.0 ENGINEERED SAFETY TEATURES i
6.10 It is our understanding that all current B&W designs call for core barrel vent valves.
Provide a detailed description and evaluation of your core barrel valves.
Reference to an existing design on an earlier plant (e.g. Oconec, if applicable). Which tests and analyses done for other vent valve designs are applicable to the WNP-1 design?
If vent valves are not to be used picase 3
justify.
6.11 Provide or reference an analysis for CFT line breaks including the effects from a small break up to and including a guillotine break of the CFT injection line.
6.12 Referring to Figure 6.3-1, ECCS P&ID, note the piping classi-fication (602) specified for the LPI/DHR 12-in. injection lines.
If either check valve CFS V-11B or CFS V-11C were to leak, parts of these injection lines could be subjected to RCS pressure.
Justify the use of pipe class 602 in light of the fact a.
that a singic check valve failure could subject this pipe to RCS pressure. Note that Table 6.3-3 rates some of this piping for 2500 psig at 350'F.
b.
Specify any surveillance for leakage provided in the injection line between check valves CFS V-11 and DHR V-37.
6.13 It is our position that the normally closed valves in each of two lines between the HP pump suction header and the DH pump discharge lines are not acceptable as proposed. These two valves should be remote manual valves with indication and control from the control room to facilitate system align-ment as required for intermediate or small breaks.
6.14 In both the Bellefonte and Greenwood plants the DHR/LPI pump discharge valves (DHR V-34 A&B) are normally open. According to Figure 6.3-1 they are normally closed on WNP-1 and there-fore must be opened for proper operation of the low pressure injection system. We believe the WNP-1 design is a less reliable arrangement. Justify why these valves are kept normally closed.
6.15 The ECCS is required to be designed for all break sizes from small leaks within the capability of the makeup system up to the doubic ended rupture of the largest pipe.
Provide or ref-crence small break analyses applicable to the WNP-1 reactor.
6.16 List in one convenient tabic the minimum number of components, as discussed in Section 6.3.2.2, that are required to operate for all ranges of reactor coolant system break sizes,and the number of components availabic.
6.17 Section 6.3.2.2.2 does not state that the DH and HP pump seals
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and other components will be designed for radioactivity that could be present for the recirculation cooling mode.
Correct this deficiency.
6.18 In Section 6.3.2.7, it is stated that a curve showing the NPSH requirements for the DHR/LPI pumps as a function of flow will be provided during construction permit review. Please supply this curve or state on what date it will be provided.
6.19 State the effect on LPI operation if the DHR pump test line valves to the BHST are inadvertantly left open following a test and a subsequent LOCA calls for LPI operation.
Specify the method used to give assurance that these valves are closed following a test.
Give the basis used in selecting the test line size.
6.20 Refer to Section 6.3.2.17, Manual Actions.
Specify in complete detail the type and location of information available to the operator and all actions (e.g., push button, read meter, etc.)
required of the operator to accomplish (a) the switching of ECC suction from the BUST to the sump and (b) the alignment of HP and DH systems for high pressure recirculation.
Include in your discussion the response to request 6.13 above.
The infermatien cupplied ir. E.0.2.17 is c.ot non sumplete enuugh k3 cunstruce the complete scenario of tnese actions or to judge the reliability of the proposed systen.
6.21 Section 6.3.4 should be amended to include periodic testing of the capability to realign pump suction and discharge by remote valve i
operations from the control room (refer to Response 6.13, and 6.20 above).
6.22 Explain how the testing requirements for the high pressure injection system (Section 6.3.4.2) conform with Regulatory Guide 1.68, Appendix A.9.a.
If your periodic testing is done with the RC system pressure at 600 psig, state how the pump performance will be related to its required performance under accident conditions.
6.23 On Figure 3.2-3 valves DHR-V78 A&B (DHR pump suction valves from SUST) are shown normally closed. On Figure 3.2-4 and Table 6.3-9, they are specified normally open with ESFAS actuation.
In Table 7.3-3 they are not listed as being actuated by the ESFAS.
Please state whether these two valves are normally open or closed and whether they are opened by an ESFAS signal.
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i 6.24 The following question assumes core barret vent valves are to be included in the '-lt;P-1 design (see question 6.10).
How were the core barrel vent valves treated in the thermal hydraulic analysis i
for normal operation? What assunptions were nade in this regard for the analyses of cccidents and transients?
It is our position that one valve less than the minimum detectable number of stuck open vent valves should be assumed to be open in the analyses for the thermal hydraulic design of the reactor coolant system and core and for all transients. 'Inat is that minimum number?
How is detection acccmplished?
6.25 In figures 3.2-3 and 6.3-) cross-connections netween the low pressure injection piping including check valves and flow elements are sho.n.
Provide drawings, an31ysis, and design bases for these flow elements.
How will the as-built ficw split performance of the cross-connections be evaluated during pre-operational testing?
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15.0 ACCID MT AtlALYSES 15.4 It is not clear from the information presented in Chapter 15 which equipment is assumed to function for each event analyzed. Thus, for example, it is impossible to determine how the single failure criterion was applied, if at all, for any event, except the LOCA and secondary system break accident.
Provide or reference a table for each event analyzed to show which equipment is assumed to oper-ate and which is assumed to be inoperative.
Include the assumptions for offsite and onsite power sources.
For each event for which the single failure criterion or other measure of reliability margin has not been incorporated in the design, provide an assessment of the sensitivity of the plant response to the assumed level of perfor-mance of the safety or auxiliary systems.
15.5 Describe in detail the method of computing the maximum dilution flow rate of 200 gpn given in Table 15.1.41.
Include and justify i
the assumed flow split between RCS makeup and reactor pump seals.
Specify the assumed number of operating makeup pumps and relate that number to the maximum possible number under normal plant operations.
15.6 In various Sections of Chapter 15 BSW toafcal recort RA'J-10074 is refercaced; i.e., h : tion: 15.1.2.2.2,15.1.1.2.2,15.1.4.2.2, 15.1.5.2.2, 15.1.6.2.2, 15.1.8.2.2, etc.
The reference should apply to the CADD computer code which is reported in BAW-10076.
Please j
make the appropriate corrections.
15.7 Ocction 15.1.9.2.5 lists the sequence of events assumed to occur following a station blackout.
Event "a" states that there will be further opening of the atmospheric dump valves 10 minutes after the loss of power.
List which atmospheric dump valves and safety valves are assumed to open following a station blackout, when they are opened, how long they are assumed to stay open, and by what action the.y are opened (automatic or manual).
State why this particular valve operating sequence is assumed to result in the largest expected radiological hazard from this transient.
15.8 Explain why a less negative Doppler coefficient (-1.42x10-5 4gjgjp) was used to evaluate the excessive heat removal accident (Section 15.1.10) 5 4K/K/F).then was used to evaluate the steam l
(-1.7x10-15.9 Table 6.2-27 states that the main steam isolation valves fail open upon loss of power.
The steam line break analysis of Section 15.1.14 assumes that one steam generator can be repressuri:ed with closure of the ftSIV's following a steam line break.
With this in mind state any differences in the results of your steam line break analysis assumingeither(1) concurrent loss of all offsite power or (2) con-current turbine trip.
15.10 Provide information to demonstrate that an 11.17 ft2 pump suction break or split does not yield peak clad temperatures in excess of thosegiveninBAW-10065andDAW-10065, Supplement 1(asapplicable; j
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see question 6.10).
15.11 In Figure 15.1.14-1, the feedwater flow is shown constant at 100%
for the first 17 seconos fol1cwing the steam line break. State what assumptions about the feedwater control system's reaction to the break were made to arrive at this curve and why these as-sumptions give the most conservative estimate of the minimum subcritical margin, mass and energy release to the containment building.
15.12 What system transient analysis method was used for the Steam Generator Tube Rupture accident? Reference or describe.
15.13 What instrumentation would be relied on to single out steam gen-erator tube failure as the cause of an event so that the reactor operator would know that the required action at 15 minutes must be accomplished (see table 15.1.17-2)? Our concern is that a number of other possible events, e.g., a small pipe break LOCA for which no operator action is required, would be incorrectly diagnosed by the ocerator. The operator could then fail to achieve the proper manual action at 15 minutes.
15.14 What rational or analysis was used to show that steam venting to the atmosonore from the affer.ted 5tm. r-wr.elor unn* '4 minutes cf tc. a sta = genarstor tuba ruptu?c (acc T:blil5.1.17-0) and from the unaffccted steam gdnerator 108 minutes after the rupture?
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