ML19220A721
| ML19220A721 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/19/1974 |
| From: | Stello V US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Moore V Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904240173 | |
| Download: ML19220A721 (6) | |
Text
,.
A"^.
Docket i STN-50-320 j 3
Yoss.A. Moore, Jr., Assistant Director for LWR's, Group 2, L F!RST ROUND QUESTIONS - TH2EE MILE ISLAND NUCLEAR STATION, UNIT 2 Plant Name: Three Mile Island, Unit 2 Licensing Stage: OL Docket No: STN 50-320 Responsible Branch and Project Manager: LWR 2-2, B. Washburn Technical Review Branch Involved: Reacter Systems Branch Requested Cocpletion Date: August 2L 1974 Description of Review: First Round Questions Review Status: Awaiting Information Adequate responses to the enclosed list of questions and com:nts are required before we can complete our review of the subject appli-cation. Cocnit:nents with respect to 10 CRF Part 50.46, Appendix X and WASH 1270 are required from the applicant.
These questions are the result of the review by the Reactor Systems Branch of sections 4.4, 5.1, 5.2.2, 5.3, 5.5, 6.3,15 and 16 of the SAR. We will have further questions with regard to the Technical Specifications (Chapter 16) and testing of the ECCS (Chapter 14, which is incomplete).
o.w.nl sined da }
Tietcr Stello #
Victor Stello, Jr., Assistant Director for Reactor Safety Directorate of Licensing
/
Enclosure:
Questions ce; S. Hanaue;, DRTA F. Schroeder, L A. Giarbusso, L c3 173 W. Mcdonald, L w/o encl.
K. Kniel, L B. Washburn, L
> O--
N C l73 ~. %:l\\
T. Novak, L
/,
/
J. Watt, L
, /'.
u m.
R. Kir M [ L L:RSB
,L L:RS
'"--~ g,- --3, L:RSS L:R
~ -V f-W.~ Hodges,-~ L
~
T-
. S. Varga,.L JWa.tt:ch W aners.
60vak.
.VStello au - - =
- 7
= = + -
_SLt[L74_._
8// 5'/74 8/// /74 8/
/7d Form AEC-518 (Rev. 9-53) AICM 0240 2 v. s. eovs==es amt emwvme orrec as e.74.sae.s ee
u
.~ w THREE MILE ISLAND 21.18 Provide a schedule for the submittal of a review of (4.3.1.3) the shutdcwn system design, plans for any oroposed plant changes required to make the consequences of an anticipated transient without scr:m acceptable and the results of supporting analysis as required by paragraph II-3 of appendix A to
'<! ASH-1270.
21.19 The thermal-hydraulic design basis requires that the minimum DNOR under operating canditions and transients does not fall below 1.3.
At a given value of maximum linear heat generation rate, the radial po'.ler ra tio affects DNER more than the axial power ratio.
There-fore, merely specifying the maximum linear heat gen-eration rate and the pecduct of axial and radial peaking values coes not guarantee that the minimum DNBR limit will not be exceeded.
'Jhat assurance is there that the racial ceakinc factor will not exceed i
~
the values listed in Saction 4.4 and the values used l
for the safety evaluation of the plant in Section i
15.
21.20 Table 4.4-1 lists Rancho Seco as an essentially identical (4.4.2.1)
NSSS. Rancho Seco was granted a license limiting its power to 2553 M',lt subject to later review cf startup reports and initial operating ex?eriences. Also, satisfactory crerating experience of the orotctype j
Oconee Unit 1 is recuired.
Summarize this experience anc l
show how it justifies the design thermal rating of Three Mile Island Unit 2.
f 21.21 Identify reactor internal elements critical to the safe l'
(4.4.2.7) operation and control of the reactor.
Tabulate for these elements the limiting design loads along with the most severe up, down and' horizontal loads predicted
~
during transient analysis.
Identify the events creating the most severe loads.
21.22 Provide the results of the calculation of maximum fuel (4.4.3.5) clad strain for coerational transients to end of life.
^
e c + >~ n C
i
me.
1.r, 21.23 In addition to load changes at constant puno combinations.
(4.4.3.5) the reactor ccolant system must be demonstrated to be free of undamped oscillations or other hydraulic instabilities for all conditions of steady state operation, for all operational transients, for all load following maneuvers and for partial locp oceration.
Provide analysis, operational experience and experimental results proViding this assurance.
21.24 Discuss experience in observing crud or scale build-up (4.4.4)
(or absence of) during the life of a plant.
Discuss how crud build-uo is considered in heat transfer analysis and ccmponent design.
21.25 Cercribe anc discussinstrumentation for vibration and (4.4.5) loose parts monitoring in tne reactor coolant system.
21.26 Identify the margins in net positive suction nead for tne (5.2.1.1) operating main coolant pumos ahen operating with cne or two pumps shut down.
21.27 Uhat is the allowable back nressure for the safety valves?
(5.2.2.2)
What is the basis for this limit? Provide the method t
used, including exterimental verification, in deternining that the back pressure limit is not exceede.
If tnis limit is exceeded, what would be the effect on safety valve relief capacity?
21.28 Show that all the assumptions and initial conditions used (5.2.2.3) in the BAU-10043 analysis are applicable to the Three Mile Island # 2 plant.
i I
21.29 BA'J-10043 does not provide the basic plant parameter such i
('5. 2. 2. 3 )
as plant geccetry and pcwer level.
Further, BA'!-lC0a3 c
does not provide the set points for both tne primary and j
the secondary safety valves.
Provide this information.
i 21.30 BA'd-10043 does not address the severity of a comolete loss (5.2.2.3) of feedwater on the overpre:sure protection cacacity provided.
Provide the analysis to substantiate the adequacy of safety i
valve discharge capaciti'es for a ccmplete loss of feecwater transient.
.=
L,.,
,_ 6 t)
g r
- s. _r.-
21.31 In the EA'.1-lC043 analysis, pressurizer scray is assumed (5.2.2.3) not to operate although the hign pressurizer pressure signal was used to scrca the reactor.
Provide an analysis where the spray is assumed to operate and therefore scram is delayed.
21.32 Shcw that the pressuri:er does not go solid for any over-(5.2.2.3) pressure transients.
Otherwise, provide the bases for the Water discharge rates through the safety valves.
21.33 The capability of the RHR system to perform its srutdcwn (5.5.7) cooling function assuming the most restrictive sin,1c active failure in the RHR systen has r.ot been demonstrated.
The RER system is not single failure pecof and, therefore, violates the intent of AEC General Design Criterion 34.
An example of a single failure which could render the RHR system inoperable is a failure-to-open of one of the isolation valves in the RHR line leading from its associated recirculation 1000.
Such a single failure could place the reactor in the position of not being able to achieve a cold chutd,in conditich..-ithin a reacanabic period of time.
It may be possible that scoe "bco strap" type of operation cu: side of the RHR systam coulo be effective in achieving a degree of sh0tJc.tn capability (such as with the ECCS), however, it is the intent of GDC 34 that the system normally utilized to place +ne plant in a cold shutdoun condition (the RHR systems be single failure proof.
Also, since the RHR system is a icw pressure system for "thich overpressure protection is required, any design modifications should not reduce the level of protection against overpressurizaticq.
The RHR system should be modified so as to be immune to single active failures before final Regulatory staff approval.
21.34 The AEC " Interim Acceptance Criteria" has been superseded
( 6. 3.1.1 )
as stated in the Federal Register, Vol. 39, No. 3-Friday, January 1, 1974.
It is required.for Three.' tile Island 2 that analysis and ' evaluation of ECCS cooling performance follcuing costulated loss-of-coolant accidents shall be performec in accordance with the recuirements of 10CFR 50.46 using an Evaluation Mccel in conformance with ?cpendix K.
A ccnmitment is recuired frca the acclicant icentifying when the Safety Analysis Report will be revised and re-submi tted so tha t the review may procaed. This co ment applies to Chapters 5 and 15.
["(%
)AC
'& a
.s i b
.s w
N 21.35 Refer to Figure 6.3-1, the ECCS' PAID.
Starting at either ECC (6.3.2) vessel injecticn no le, trace back along the piping tcuard the low pressure system, through two check valves, the reactor building boundary, and a normally open matcr operated valve.
At this last valve there is a transition frca high pressure to lcw pressure piping. Our cencern is that no means are provided to detect leakage frca the reactor coolant system back through the first (relative to the RCS) check valve, or from the core flooding tank (CFT) back through the second check valve.
In the latter case, a decreasing CFT level may be sufficient indication of leakage for the operator.
However, in the former case, undetected leakage from the RSC could pressurize the line between the two check valves for an undetermined period of time.
Subsequent failure of the second check valve or the CFT check valve would result in a LOCA (outside containment in the first case, inside in the second) with diminished ECCS capability.
Thus, the failure of one check valve could lead to a LOCA and a der aded ECCS. A second concern is that no pressure relief device?'
are shown on the Figure.
A change in design or monitoring should be made so that full credit can be taken for both check valves as protection against
, back leakage from the RCS.
Such a change could take the form of a pressure indicator between the check valves, use of high pressure pioing throughout, additional valves, addition of safety valves, different valve administrative alignment, or a comb.ination of these.
21.36 The ECCS design sho.;n in Figure 6.3.1 does not meet the (6.3.2) requirements of GDC 35. A failure of one injection line resulting in a LOCA coupled with a single failure in the other injection train would incapacitate the ECCS.
The proposed design for such S&'d plants as North Anna 3/4, Bellefonte, Greenwood, and 'dPPSS are examples of acceptable designs with respect to icw pressure injection.
Provide a description of the re-designed ECC system which fully complies with GCC 35.
Include a discussion of the design basis and an evaluation of the operation of the system.
21.37 Provide an enlarged (legible) Figure 10.1-2 (10.1) r, ~
4 t.*,
s-21.38 Provide a curve showing the effect of reactivity inserticn (15.1.2) rate on minimum DNSR.
Forthis insertion rate which giYes the ninimum CNER provide a curve of ;:,3R vs. time. Assume that ' the transient starts at 102% rated power, 2132 psie and 559'F inlet temperature.
21.39
'4 hat values of radial peaking factors and enthalpy rise (15.1.2) factors (F. g) were used in the analyses presented in Chapter
~
15?
21.40 For the loss of Coolant Flow Analysis, provide curves of (15.1.5)
DNSR vs time and not spot heat flux vs. time for the four pump shutdcwn.
21.41.
What is the DNBR for steady-state operation at the conditions (15.1. 5 )
assumed for the start of the ficw coast-down transient (102% rated po,eer, 2135 psia and 559 F inlet temperature)?
21.42 Provide evidence that the uncertainty in inlet temperature (15.1.5) is only 2 F.
Reference:
Table 15.1.5-1.
i 21.43 Determine if one motor driven emercency feed cump (470 GP")
(15.1.8) is sufficient to bring the plant to a safe shutdown condition.
The event pcstulated is as.folicws: A rupture occurs in the high energy steam supoly line to the emergency feedwater pump turbine. This is coupled usitn the active failure of one motor driven emergency feed pump.
Provide analysis and discussion of this postulated event.
t 21.44 Provide an analysis for a feedwater line ructure.
In the i
(15.1.8) analysis justify the methed used to calculate break fic.i, the sizes and locations of breaks.
Show that the, single i
failures considered in the analysis are the most liaiting i
ones. Further, if the pressurizer goes solid as a result of this accident, provide bases for water discharge rates through safety valves.
21.45 The set points for the various overpressure protection (15.2.2) devices should be stated under " Specifications."
Lv
.c ra