ML19209C696
| ML19209C696 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/23/1970 |
| From: | Ross D US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Long C US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 7910170863 | |
| Download: ML19209C696 (25) | |
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ATOMIC ENERGY COf/.'AISSION s
'I b')*f WASHINGTON, D.C. 20545 I'
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April 23, 1970 Cnarles G. Long, Chief, PWR Project Branch 2 Division of Reactor Licensing PRELIMINARY REVIEW OF TMI-l OPERATING LICENSE APPLICATION (DOCKET 50-289)
I have made a preliminar'/ review of TMI-l FSAR and made some notes,
and questions, which are attached.
These, with additional input f rom DRS, will be used as input to the first technical meeting.
, w - e-c h 0 *
Denwood F. Ross PWR Project Branch 2 Division of Reactor Licensing Attachment cc:
Docket File A. Schwencer D. F. Ross FWR-2 Reading 1476 082 7910170
[
PRELIMINARY REVIEW OF Tl!REE MILE ISLAND UNIT NO. 1 FSAR CHAPTER I 1.
Jersey Central Power & Light is new 25% owner.
According to the letter of transmittal, Met-Ed is still responsible for design, construction, operation and maintenance.
~.
2.
The FSAR ic asking for a licensed power level of 2535 FNt (core).
3.
The FSAR letter of transmittal states that the FSAR (termed Amendment 12) replaces all of the preceding documents (PSAR, Amendments 1-11, and letter information).
4.
Nineteen design changes (since in PSAR) are conveniently summarized in Section 1.3.2.
5.
Section 1.4 lists the 70 GDC and their application to TMI-1.
6.
Section 1.5 is a summary of R&D:
a.
OTSG - baced on BAW-10002.
b.
Control Rod Drive - based on BAW-1000/.
c.
Self-Powered In-Core Detector - based on BAW-10001.
d.
Thermal-Hydraulic - based on BAW-10012 (Vessel Model).
e.
ECCS and Vent Valves - based on BAW-10008 and BAW-10005.
f.
Fuel Rod Clad Failure - s tudies underway; no submittal date forecast for the U&W topical report (BAW-10009).
g.
Xenon Oscillations - ref _rs to B AW-10010, Parts 1 and 2.
Part 3 is yet to be filed.
h.
Iodine Removal - Refe rs to BAW-10017; a future supplement on effectiveness is promised.
7.
Fif teen pages are furnished on QA.
1476 083
e 2
CHAPTER II 1.
Some details on the probability of airplane strikes. are included in Section 2.4.2.
Since the design bases (200 knots, 200,000 lbs) were agreed on at the CP, and there is no new information that negates the basis, this aspect of the FSAR is not a major review Section 2.4.2.5 may use probability as a tool for setting area.
the design basis fire and is worth reviewing.
I. Pinkel has been engaged as a consultant.
2.
Two years of onsite meteorological data are analyzed. Based on probability, an accident X/Q value of 2 x 10~ sec/m is proposed for the 2 -hour dose at the site boundary.
The probability that this figure would be exceeded is 57., using 2 years of data.
k'e had a report at the CP stage from ESSA to the effect that 5 x 10~
sec/m should be used at the exclusion distance.
Since this new X/Qvaluereducescalculatedaccidentdosesbytheratiof,we should expect a meteorology rpview to be a significant review item.
3.
The hydrology section, ficod studies, lists (on page 2-26) a design 6
flood of 1.1 x 10 cfs.
A letter from Met-Ed to DRL dated July 3, 6
1969 states that the TMI dike was designed to accommodate 1.1 x 10 cfs, but that they understand the maximum Corps of Engineers' calcu-6 lation may increase to 1.6-1.7 x 10 cfs.
They (Met-Ed) intend to provide safe shutdown using the higher value (according to the letter).
However the FSAR letter of trans :.ittal (dated March 2, 1970) replaced all previous commitments. Thus the ability to cope 6
with the PMF (1.6-1.7 x 10 cfs?) must be a major review item.
The July 3, 1969 commitment is res tated on page 2-30.
k'e need to find out what that commitment really means.
}476 084
\\
3 4.
Sectica 2.6.5 discusses the ability of features associated with flood protection to survive the PMF.
These should be reviewed 6
thoroughly, especially if there is partial inundat. ion, at 1.6 x 10 cfs, and consequent erosion.
The access bridge is apparently 6
6 designed to 1.1 x 10 cfs; what happens at 1.6 x 10 cfs?
1476 085
4 CHAPTER III 1.
Clad strain limits are:
(a)
Stresses such as circumferential membrane stress should not exceed yield, or 75% of stress rupture life.
(b)
Clad plastic strain due to fuel swelling, the rmal ratcheting and creep, ar.d internal gas pressure limited to about 1%.
(Page 3-5, Volume 1)
Are these two statements contradictory?
2.
On pages 3-18 and 3-19 (Volume 1) it appears that the vessel model flow tests did supply input to thermal-hydraulic design. We had been told at a meeting, by B&W, that this was not going to be done.
3.
Why wouldn't flux shapes occur that are tilted towards the top half of ine core?
4.
Where 'does Met-Ed stand on BAW-10000 (a new report on heat trans fe r) ?
T476 08o6
5 CHAPTER IV REACTOR C00LWT SYSTE:1 1.
The relief capacity of each code safety valve is about 312,000 lb/hr at 2500 psi. What is the corresponding figure for liquid cf flux?
2.
Is it practical to set low pressure alarm at 1320 psia and low pressure scram at 1800 psi?
(That is, is the equipment of such narrow range that you can read the dif ference between 1800 and 1820?)
3.
Regarding Section 4.1.2.5 Seismic Lcads and LOCA loads: To what extent do the vessel and internals dif fer f rom Oconee in physical construction?
4.
What is the design basis of the vessel cavity? How long does it take to fill with water following a LOCA through the drain line
'from the refueling cienal? Is this water stagnant?
5.
Is a continuous boron inventory accomplished (relative to " hideout")?
What limits do you plan on unexplained reactivity changes?
6.
What RCS motor parameters (such as loss of cooling, bearing tempera-ture, winding temperature, speed, low oil, voltage / frequency), can shut the motor down automatically?
7.
If the flow is 5% greater than the design basis (which appears reasonable on the basis of Figure 4-8) and core LP is thereby 10%
greater, what would be the ef fect of this additional loading insof ar as normal and/or accident modes?
8.
Will vessel head temperatures be monitored, and (if so) hcw will they be used?
1 A76 087
6 9.
Has the (one or more) primary pump been previcusly operated in a calibration leop?
h' hat about the flow tube calibration?
(Provide details.)
10.
k'h a t is the gal / inch of the RB su=p, and what is the span (inches and gallons) f rom sump pump (s) on to of f ?
11.
Does the pressurizer electromatic relief valve work en emergency power?
12.
Describe the low-level interlock on the pressurizer heaters.
13.
As an example, show the calculations on the design of the pressurizer foundation and supports, in regard to the combined earthquake and circumferential rupture of the 10-inch surge line (that is, show that the pressurizer remains tied down).
14.
Describe the missile protection design procedures used in es tablish-ing the thicknesses of the special missile shields (page 4-21).
Stipulate typical missiles and how concrete penetration is calculated.
15.
At the top of page 4-30 it is stated (and refers to 14.2) that if the equivalent of one code safety valve sticks open, one HPCI pump can easily protect the core.
Comment on this analysis :
One code safety valve passes 320,000 lb/hr (which is about 640 gpn of makeup water) at relief pressure of 2500 psi.
This seems quite a bit higher than the HP pump capacity at 2500 psi.
Please review the FSAR statement, or comment on mine.
j476 088
7 16.
Ques tion on Section 4. 3.10. 3:
Describe the routine measurements or observations made on the makeup tank level, including the estimated precision.
How is a water inventory maintained for compliance with leak-rate tech specs.
l'7 '. Can other-than-borated water ever be in the DH loops?
If not, how is crystallization avoided?
If so, is there a potential for a fresh water accident.
18.
Regarding Section 4.3.11.1:
Which organization did each dynamic analysis of piping systems? Cite references. Furnish details of separate review of stress analyses.
19.
Section 4.4.1 treats in-service inspection capabilities.
State your proposed procedures, including tech spec sections. What weld history data are retained by Met-Ed?
20.
How will dissimilar welds on CRD's be inspected routinely?
21.
Is full flow (and temperature) established without a core (in functional testing)?
i 1476 089
Describe the sump pump and level alarm features for the DH pump system, including emergency pcwer arrangement.
Same comment for spray pump.
2.
Elaborate on the last sentence in the second paragraph of p 6-3 regarding leakage from the recirculation line guard pipe.
3.
The makeup tank holds several thousand gallons.
Presumably this eenld be pure water, if you are in a deborating cycle. On an ESF signal, would all of this water be injected before any BWST water went in? What would be the nuclear consequences?
(Cold, tresh slug of water?)
4.
The moderator dilution accident is not fully described in terms of methods of analysis, peak pcwer reached, power and flow terminating mechanics, mixing justification. Would the single operator error of starting a HP pump cause this event?
5.
How will the analysis of accumulator flow rates vs time be verified?
6.
Is level in the R3 sump monitored and is that instrumentation designed to survive a LOCA radiation environment?
7.
How does the minimum temperature in a CF tank relate to the saturation temperature for the ma::imum boron concentration?
1476 090
9 8.
Describe the procedures for isolating one or both Ci tanks during routine depressurization. Will CF tanks be routinely purged?
9.
Please make available Reference 1 of Chapter 6 regarding qualifica-tion of DIl pump seal under temperature and pressure. Was radiation considered?
(10 to 10 r expected during course of accident?)
10.
According to Figure 6-6, the DH pump has 350 ft TDH at 3000 gpm, and requires about 14 f t NPSH.
If measured flow rate is 10" higher with VWO, NPSH requirement increases 3-4 feet. Will the DH flow be throttled so as to not exceed 3000 gpm?
11.
What limits will apply to measured leak rates in the DH heat exchangers?
12.
What temperature margin (above boron crystallization) is maintained in the BWST?
Is circulation accomplished?
13.
What keeps the DH pumps f rom injecting NaOH? Or is this a design?
14.
For nhat size breaks would the building coolers work to keep RB pressure below 30 psig? Have dose consequences been evaluated for those events (as no chemical spray additive would be in evidence)?
15.
Refer to Table 6-9: Under normal condition the cooler flow rate is 430 gpm and the temperature rise is 11* F.
This represents a power of about 700 kW.
If there is a 1 gpm primary leak rate, the energy released in going to ambient condition is about 500 Btu /lb, and at 1 gpm represents about 20 kW.
Is it contended that the 20 kW increase can be measured above a residual of 700 kW (as a leak de te cto r) ?
16.
Refer to Table 6-10: Tne leakage from ESF equipment is about 2-1/2 gal /hr. Has heat exchanger leakage been considered?
1 A76 00
10 17.
It appears f rom Figure 6-4 that particulate (roughing) filters are Provided in the RB air recirculation.
If so, describe them, their ability to survive the LOCA, and their function (if any) in reducin;;
the dose consequences.
9
/
e 1476 092
11 CHAPTER VII OW 0lO I I INSTRUME::TATIO:;
l][] ] A, [I
?Y T'41 fil
-u 1.
Will a neutron detector really survive an MllE--especially suspended from a long pole?
2.
Has BAW-10003 been made available yet?
(Qualification Tes ting of Protection System Instrumentation).
3.
What procedures are envisioned for regrouping of CRD's?
4.
Where is the system dispatcher? Describe his duties, or abilities,
relative to load control of the unit.
5.
Why is a 7-decade source-range instrument used? Where else has it baan used, and what testa have been performed to demenctrate itc linearity? Where is count rate for typical startup (lower limit)?
6.
Describe your plans for setting the compensating voltage for the intermediate channels, both for a cold start and hot restart.
/
7.
Is it indicated to the operator that the source range high voltage is on (since various interlocks can automatically turn it off)?
8.
Since there are only two outlet temperature sensors per hot leg, how is IEEE met for 1-loop operation?
(Possibly same comment for flow rate detection.) Since you do not propose 1-loop operation, does 1 pump each loop negate any safety instrumentation?
9.
Provide more informction about the range of the pressurizer level instrumentation, and switching (if necessary).
1476 093
12 10.
What is the purpose of the startup reactor coolant pump interlock (with inlet temperature)? Could this trip one pump if inlet te'mperature drops?
11.
Describe the sealing mechanism (to the vessel) for the in-core detecter, and the method by which leaks would be detected.
1476 094 1
13 CHAPTER 9 REVIEG AUXILIARY AND EMERGE::CY SYSTE'!S 1.
H5ve single f ailt.res in the makeup and purification system been identified that would lead to excessive makeup (overfill) or deficient makeup?
2'.
How long could the control rod drives function without the cooling coil, operable on the intermediate cooling system?' Is one of the 70 CDC violated?
3.
In regard to the heat removal capacity of the spent fuel coolers for storage of 1-1/3 cores:
What is the delay time for unloading the core?
a.
b.
Does each spent fuel cooler transfer 25.85 MWt during this condition? If so, and if cooling water is 95 F (Table 9-8) and pool water is 126 F (p 9-18) then consider:
l From__ Q V
To 2
Pool a
as a I
Pool
\\
I^ 'v-U~V
~t F = 1000 gpm l
V 6
]
f
'2 t y NSW Cooling Water 1476 09c3
14 T = 126* F (p 9-18) y t = 95* F (p 9-53, or Table 9-8) 1 6
Since P = 25.85 x 10 Btu /hr = 7570 kW F = 1000 gpm P = 0.147 F
- AT P in kW AT = 7570 147 F in gpm AT in *F AT = 51.5* F T2 = 126-51.5 = 74.5* F Obticualy T cannot 've less than i.1, or 95" F.
2 I suggest that T = 126' F 2
T = 126 + 51.5 = 177.5* F y
AT In this case the efficiency n = 177.!-95
,31.5 82.5 n = 62,';; still high 6
For the case where Q = 8.75 x 10 Btu /br = 2560 kW 2560 T2 " 147 = 17.5* F 1476 096
O 15 T = 120-17.5 = 102.5 y
n = yf ; 5 j
70% (very high)
=
=
If n = 0.4, T =95+1'f=95+44=139*F,not120'F.
1 0.9 4.
Suppose that for long-term fuel storage, Pool 3 is -isolated from
- Pool A by the water gate, and that Pool A is drained (p 9-19).
What is the seismic design basis of the water gate?
5.
How big is the line that transfers water from the reactor vessel area during refueling back to the B'4ST?
6.
Reference Figure 9-10: Clarify the specification symbols on the 12-inch line connecting the RCS hot leg to the DH suction header.
Where is the transition from 2500 psi to 300 -si design basis?
How are valves DH-V1, V2 administratively controlled? What is the design basis of DH-V377 1476 097
16 CHAPTER X TMI (il Calculate secondary activity inventory of the steam generator.
Provide assumptions on primary activity and inleakage, blowdown, carryout due to entrained moisture, partitioning, decay.
2,.
Does the reactor trip (now or later) on turbine trip (with or without offsite pcwer)?
3.
Is there a load dispatch control from remote stations?
4.
How do you open COV-1, Condensate booster Pump bypass valve.
5.
The Powdex syt :em (condensate treatment) is Class III.
How would the plant be brought to cold shutdown if this system were blocked?
6.
Indicate the Class I-III interf aces on Figure 10-2.
7.
As an illustration, state the valves that must open for emergency feedwater pump EFP-2A to supply water to steam generator lA.
What valves stay open, and, for those with operators, what causes remote op_ning?
8.
What is the lower operating limit of the turbine-driven emergency feedwater pump (steam requirements vs water delivered)?
'476 098
17 9.
What is the basis for regulating flow rate in the emergency feed-kater chain?
- 10..
How is bypass valve FW-V6 opened and what dictates its use?
11'.
Does high condenser vacuum trip the main FW pumps, and does this open FW-V6?
12.
What s tarts emergency FW pumps (elaborate en page 10-3, third paragrapF)?
13.
How vulnerable (seismic, aircraf t) are suction lines to emergency FW pumps?
14.
What is the closing time of MSIV 1A, B, C, D? What is the closing basis?
15.
What opens MSV 2A, or 23 and MSV 10A, 10B, admitting steam to emergency FW pump?
16.
Does MSV 8A or 8B close when EF-P1 comes on? How powered?
4S-I 17.
On Figure 10-3, does denote a seismic class 4 g _ 777 change? If so clarify Figure 10-3, especially around main s team lines and main feedwater pumps.
1A76 099 e
w D % l5) f[DlGflf in kuu)di]d@C"A 18 u
18.
Reference page 10-4, last paragraph:
Are MSV-13A, 13B the two valves that "saap" open to bring FW pump up to governor speed?
19'.
- Confirm this automatic sequence on turbine trip:
a.
Turbine stcp valves clo.;e
.b.
Main FW pumps trip c.
Condensate and condensate booster pumps trip d.
Small steam supply valve opens to bring emergency FW pump to governor speed.
Then, manually, operator opens large s tdan admission valve (10A, B?)
(I thought everything was automatic.)
What i f turbine-driv an punp dooso't funct on? Any entemetic action i
on electric pumps?
20.
Describe the condenser of f-gas monitor and the basis for setting the alarm. How will released activity be considered in annual release limit calculations?
21.
Describe the electrical. connection that provides reactor trip upon loss of condensate pumps (or is it high reactor pressure?).
22.
Following turbine trip with no loss of offsite power, will the reactor be lef t at some " idling" level? If so, what are the minimum levels in hotwell or CST such that procedures for cold shutdown will be implemented?
Is it the intent that the emergency river source obviates the need for such minima?
1476 100
19 23.
Provide more on tests and inspections relative to emergency feed-water sys tems.
~.
t 1476 101 i
e
20 CHAPTER 11 REVIEW TMI #1 1.,
All of your referenc'.s on escape rate coefficients are at least 5 years old.
To what extent do they apply to the B&W design f uel element operating ~18 kP/ft; Zr-clad; 40,000 mwd /MTU.
2.
Does PMF flood any radwaste components or systems?
3.
Regarding 5000 gpm of cooling tower purge (page 11-5):
Is this maintained even if station is shut down?
Is flow monitored, and is there a low-flow alarm?
4.
Describe the design and calibration procedures for liquid and gas radiation monitors.
In particular show by analysis that asserted lower limits of sensitivity can be achieved.
i S.
Discuss the procedures to be used in isotopic analyses for disposing of was tes > 10~ pc/cc (refer to Table 11-7, p 11-38, activities for S r, Y, Mo, I, Cs).
6.
Since there is a mi::cd activity release, how can an activity monitor be properly calibrated unless a single.'2C is assumed?
7.
How do radiation monitors terminate an excessive liquid release?
1476 102 e memme
21 8.
Comment on this calculaticn:
Annual release = 0.05 C1 (FSAR p 11-8)
~.
Annual dilution = 5,000 E1 x 500,000 "i" x 4,000 EE min yr gcl 13
= 10 cc/yr 4
5x10 uCi
-9 Concentration
= 5 x 10 pC1/cc
=
13 10 cc or about 5% of !TC, if FTC = 10- pCi/cc (you have 0.025%, x 200 difference).
Note:.In December 1969, Connecticut Yankee released 0.5 C1 -
liquid radwaste, pluse730 Ci tritium.
In January-February 1969 Indian Point-1 reported 13-16% STC liquid release.
San Onofre had a maximum; on the period of January 1969 through June 1969 of 25% MPC, and released a total of 0.32 C1.
Can you relate this recent experience to your higher-powered plant by citing. design dif ferences in radwas te sys tem?
1A76 103
22 9.
What portions of the radwaste system are on emergency power?
10.
Clarify the control room cose figure on p 11-14 and paragraph 11.3.2.4.
Is this 3 Rem /90 days (accident) for continuous occupancy?
t 11.
The control room dose figure of.0.5 mr/hr yields an annual worker dose of 1 Rem /yr (from failed fuel activity).
Same for office, turbine building.
This seems unnecessary.
12.
As an example, describe the means by which the control rocm air is
-11 sampled for I-131 activity down to 10 pCi/cc.
'13.
What is the basis for the sensitivity setpoints on RM-L6 and L-8 (namely, no lower than 2 x 10- pCi/ cc). How would they know what isotope is being passed?
14 Reference to p 11-21, paragraph 11.5.2:
Specify the equivalent of a film badge.
15.
Summarize the results of the environmental monitoring to date.
)h
23 REVIEW OF CHAPTER 12 TMI #1 1.
Describe the duties of the station superintendent relative to Unit No. 2.
Will the superintendent have an assis tant?
2 '.
Describe the Unit 1-Unit 2 isolation procedures to be used while Unit 1 is running, and Unit 2 is being built.
3.
Will you use the B&W reactor simulator?
1476 103
24 REVIE*,i 0F CIL\\PTER 14 TMI !!1 1.
In regard to the environmental consequences of a Steam Line
' - Failure, calculate the secondary activity levels prior to the b re ak.
E:-: plain assumptions en pricary activity, primary-secondary carryover, secondary removal routes, etc.
4 2.
Explain the cethod used to calcuiate 435 gpm leak rate from one OSTG tube failure.
3.
Discuss hydrogen purging.
1476 106
.