ML20040A056

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Forwards Graph & Drafts of Second Round Questions & Analysis
ML20040A056
Person / Time
Site: 05000561
Issue date: 09/28/1976
From: Cox T
Office of Nuclear Reactor Regulation
To: Happell J
BABCOCK & WILCOX CO.
Shared Package
ML111090060 List: ... further results
References
FOIA-80-515, FOIA-80-555 NUDOCS 8201200272
Download: ML20040A056 (10)
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U.S. NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 301-492-7000 1

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B-SAR-205 SECOND ROUND QUESTIONS e24,2 o

$29)C7at S'/ STEMS 212.222 With regard to the response to question 212.35, the staff posi-(5.2.2) tion is that adequate means must exist to preclude an inadvertant overpressurization event caused by a single active component Q

failure or single operator error. Unless B&W can show that the current design would prevent exceeding the pressure-temperature operating limitations due to an overpressure event, design modifications must be proposed to provide this assurance.

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T With regard to the response to question 212.97, B&W stKra;

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212.223 (5.2.2) that for analyses of events involving upset conditions it is x

reasonable to assume the "most probable" initial values for

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operating parameters.

1, The staff position is that the events required in Chapter 15

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are not limited to realistic analyses. The Standard Review Plan j

requires that the valuas of parameters used be suitably I

conservative. For example, page 15.1.1-3 of the Standard Review Plan instructs the reviewer to compare values of parameters assumed in the analysis to the Technical Specifications.

Nominal values are not suitably conservative. In addition to B&W's conservative practice of placing trip setpoints at maximum tolerances to reflect setpoint uncertainties, we will require a selection of initial parameters to reflect the extremes of the ranges allowed by Technical Specifications, including instrument uncertainties.

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212.224 Wi':h regard to question 212.109, the response is partially (5.5.7) acceptable. Discuss the capability to bring the plant to a cold shutdown condition from full power operation using only j

. safety grade systems. Provide a protection sequence diagram similar to the FMEA figures in the response to question 212.1.

Of the 13 steps shown in Section 5.1 which are followed by

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the operator during a shutdown operation, state which ones ars essential to achieving and maintaining the plant in a cold shutdown condition from full power operation. Similarly, N

state each function in Table 5.1-10 which is essential to j

- achieving and maintaining a cold shutdown.

212.225 The' response to question 212.116 states that closure of an (5.5.7)T RHR suction valve would not necessarily cause pump damage.

Justify this contention. Some flow is indicated to be main-i tained in the pump recirculation line. Where is this cooling

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s water coming from (i.e., pump suction valve is closed) and l

1-how long would it last?

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, 212~226 With iegard to the response to question 212.127,

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(6.3.2) 1.

Relief valves DH-RV5A, -5B: Submit a plot showing the actual

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pressure transient in the low pressure piping for the worst case (i.e., HPI actuation). Show the time at which the peak pressure is attained and the times at which the DH r

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relief valves are actuated. Indicate the affect, if any, of the suction isolation valves on the peak pressure.

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i The response indicates that the + 10 psig relief valve setpoint uncertainty was not accounted for in the relief l

valve sizing. The staff position is that such uncertainties must be considered in the sizing and setpoints of all over-pressure protection devices.

2.

Relief valves CF-RVlA, -1B: A nitrogen overfill situation j

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was not addressed. Also, pressure relief should account for setpoint uncertainty.

212.227 With regard to question 212.141, the response is insufficient I

(6.3.2) to allow an adequate evaluation. Provide the safety design basis

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's of these recirculation lines.

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212.228 With regard to the response to question 212.147, the low seal (6.3.2) injection trip should be included in Table 7.3-2 as an ESFAS trip parameter.

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Also, the staff position with regard to pipe breaks is that credit i

for operator action is not allowed until 30 minutes after the first alarm. In addition, operator action outside the control i

room to mitigate the consequences of pipe breaks is not permitted.

It appears from the evaluation that the operator must; 1.

Isolate the break (from the control room).

2.

Place the spare makeup /HPI pump in operation ( outside the control room).

3.

Isolate letdown (from the control room).

While the desirability of isolating letdown is obvious, it is not clear that this action is essential. Please comment. Also, j

discuss the feasibility of adopting the capability of placing the spare pump in operation from the control room. Finally, reassess the plants' response to this event not taking credit for operator I

action before 30 minutes.

Iq 212.229 Eith regard to question 212.1, the response is partially (15.1) acceptable. Provide the following additional information:

1.

Figure 15C.13-2: " Reactor Trip" should be replaced by "CRDM."

2.

Figure 15C.13-3: "CRDM Trip" should be shown as "CRDM."

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3.

Provide a protection sequence diagram for the event described in question 212.147 since it is important to depict the different protection sequences required for makeup line breaks.

4.

Provide a protection sequence diagram for a CFT line break since it is important to depict the different protection sequences required for this type of LOCA.

5.

Submit the protection sequence diagram requested in question 212.153 for HPI line breaks.

6.

Figure 15C.14-1: The need for Secondary Steam Dump Valves disagrees with Table 15.1-4 and should be corrected.

7.

Provide e protection sequence diagram for the event analyzed in question 212.169.

8.

All protection sequence diagrams should again be reviewed by B&W for consistency with Chapter 15 analyses. For example, Figure 15C.4 is unacceptable since this event during re-fueling (Mode 4) 'is analyzed in Chapter 15 and should be reflected by a different protection sequence than shown in this subritted diagram.

9.

All events must be reviewed by B&W for applicability to the four operating modes in Table 15C-2.

Provide a summary table showing (for each mode) the applicability of the event.

If an event is not applicable for any given mode, a brief explanation should be provided. It would be expected that each event could occur in all operating modes, with few exceptions.

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212.230 The response to question 212.169 with regard to the worst-case (15.1) overpressure transient is insufficient to allow an adequate evaluation. Show that the stated tire available for operator action (before exceeding the pressure-temperature limits) would also be available at cooler reactor coolant temperatures.

Provide a discussion with regard to a normal startup (or cooldown) and the expected pressure-te.sperature combinations; special emphasis should be given to the typical high pressure-low temperature conditions normally observed during a startup or cooldown. While it is recognized that one combination of initial pressure and temperature could produce a faster and higher pressure rise, it would appear that a slower pressure rise could be worst-case at the cooler temperatures (in terms of time available before the pressure-temperature limits are exceeded).

On a P-T plot, show the initial and final conditions of each P-T combination considered. Also, provide the makeup flow rate assumed with a justification for its value.

212.231 The response to question 212.172 did not answer whether any (15.1)

Chapter 15 event would be more severe for the 3600 MWt design than for the 3800 MWt design. Therefore, unless it can be demonstrated that the most limiting transients and accidents for the 3800 MWt design would bound the 3600 MWt design, a reanalyses of Chapter 15 will be required to support licensing as a PDA for 3600 MWt.

212.232 wien regard to question 212.192, the response is insufficient to (15.1)

, allow.an adequate evaluation. Confirm that the most limiting of the events that result in a decrease in heat removal by the secondary system are the loss of feedwater with regard to primary side pressure and the turbine trip with regard to secondary side pressure. State the most limiting event with regard to core thermal margins. Provide or reference the analysis for each worst case.

212.233 With~ regard to the response to questions 212.178 through 212.182

'ertaining to the.CVCS dilution event, the staff position is that (15.1.4) p alarms shall be available during refueling from audible count rate instruce'ntation to d'etect changes in the reactivity condition of

  • the core. The analys,is of this dilution event would dictate that the operator have a prompt and definite indication of any boron dilution from the audible count rate instreuentation; therefore, for initial core loading it is prudent to require the more sensitive channels temporarily installed to provide audible indication. It is not obvious that sufficient sensitivity of the excore source range instrumentation ' exists during initial core loading to promptly alert the operator of a continuing boron dilution.

State whether the operator would have more time from the first alarm or less time if the dilution analysis during refueling assumed a higher initial boron concentration. State the concen-tration values assumed during refueling and the time predicted available after the first alarm until the system goes critical.

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t 212.234 With regard to the main steam line break, the responses to several (15.1.14) questions asked during and subsequent to the Acceptance Review show that a systems-level analyses had not been performed well enough to identify the v7rst-case steam line break (with respect to reactivity margin and DNBR). Sufficient lack of clarity continues to exist in the Section 15.1.14 discussion to warrant our position that the main steam line break analysis must be rewritten in BSAR or resubmitted in the form of a topical report. This new submittal must include the following points:

1.

A summary table which lists each break location considered (inside vs outside containment, 28" line vs 42" line, with offsite power vs without offsite power, etc). For each break the table should list the resulting reactivity margin, peak cladding temperature, and selected worst eingle active component failure.

The table should clearly show which break location is worst-case in terms of reactivity margin and which break location is worst-case in terms of ECS.www %.

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2.

For each case' represented in the above table, the text k

should present the analysis with a discussion and justifica-tion of why the selected single active component failure was considered worst-case for that break location.

3.

For the worst-case break with regard to reactivity margin and PCT, provide the time histories of Reactivity Margin, PCT, Break Flow Rate, DNBR, Pressurizer Level, Steam Generator Levels, and Steam Generator Pressures.

212.235 The description of a steam pressure regulator malfunction (Section (15.1) 15.1.36) and the inadvertent opening of a steam generator relief or safety valve (Section.15.1.37) indicated that the consequences of these events are bounded by the main steam line break. The criterion for Chapter 15 events to ensure that no fuel damage occurs is that a DNBR greater than 1.30 must be maintained throughout the transient. Since the steam line break analysis shows a DNBR less than 1.30, a reference to this analysis as a bounding calculation is not appropriate. Provide the specific analyses for the cooldown transients to show that DNBR remains greater than 1.30 for each event.

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212.236 The response to question 212.71 indicates that the rod worth (15.1) curve shown in the PSAR (Figure 15.1-1) was non-conservative.

B&W states that they will modify their scram time specs to compensate for this non-conservatism. Provide sensitivity studies to show that the limiting events in Chapter 15.0 would not become more severe with these new assumptions.

Also, the original question with regard to the cosine power shape was not answered.

In addition, for events that do not involve reactor trip, the variances in axial power shapes must be accounted for in addition to consideration of maximum design peaking factors. Show that the assumed cosine shape was conservative or resubmit analyses with the appropriate power shape of each event which does not involve reactor trip.

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221 42 (4.4.2.2) Since Equations in Hytran assume no heat generation in the fluid, justify your statement that you assume 97.3% of the heat generation occurs within the fuel rod with the remaining 2.7% generated in the coolant.

.2.W. 45 (4.4.2.2) You list three references for cladding conductivity. Since they are not all identical, detail the equation you use to calculate conductivity and explain any deviation from the references.

.22/. 44 (4.4.2.6) Justify your statement that failure to include expanded flow areas around bowed assemblies will result in less than 0.03%

error in the hot bundle flow. Detail these calculations completely and provide the basis and assumptions of your analysis.

.2.2/. 47 (4.4.2.7) In Section 4.4.2.7 correct your Section reference for a discussion of core pressure drops and hydraulic loads during accident conditions.

22/.4/,4.4-30) Why is there more than a 10% pressure drop difference (Fig.

between Greene County and BSAR-205 at the 157" level?

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(4.4.2.8.1) Pro.ide values of K in equation 4.4-7.

Give a reference c

for the values of K and indicate what experiments they c

were derived from.

.33./. 48 (4.4.2.8.4) Equation 4.4-26 has a division sign missing. Define the radiation gap conductance (hrad) for equation 4.4-28.

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-.ul.4i (4.4.2.8.5) Amendment notation was omitted in this section. Provide a schedule for completion of the Oconee 1 fuel exam and the submittal of the data reduction.

et.21. S b (16.2.3)

Provide a discussion of the role of pump monitors in preventing core DNBR from decreasing below 1.3 if there were a loss of the reactor coolant pump (s).

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