ML19321A642
| ML19321A642 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 07/21/1980 |
| From: | Moon C Office of Nuclear Reactor Regulation |
| To: | Doherty J DOHERTY, J.F. |
| References | |
| NUDOCS 8007240010 | |
| Download: ML19321A642 (38) | |
Text
e e
07/21/80 i
UNITED STATES OF AMERICA NUCLEAR REGULATORY CO MISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD i
In the Matter of
)
)
HOUSTON LIGHTING & POWER COMPANY
)
Docket No. 50-466
)
(Allens Creek Nuclear Generating
)
Station, Unit 1)
)
NRC STAFF'S PARTIAL RESPONSE TO JOHN F. DOHERTY'S THIRTEENTH SET OF INTERROGATORIES The NRC Staff responds, in part, as follows to the thirteenth set of inter-rogatories propounded by John F. Doherty in the captioned proceeding.
By agreement with Mr. Doherty, the remaining responses will be filed as soon as the necessary Staff reviewers complete current review assignments.
GENERAL INTERROGATORIES 2.
!!as the precise difference between the neutron Jynanics of large cores such as a BWR/6 as calculated by diffusion theory and that determined by neutron transport theory been deter-mined for the ACNGS core?
Response
So far as we are aware there has been no tpplication of transport theory to dynamics analysis of any light water reactor. The areas in which transport theory provides improved answers over diffusion theory (deep penetration problems, flux variations across boundaries between very dissimilar regions, etc.) are not of concern for predicting the dynamic behavior of reactors.
The use of transport theory would be expected to have a trivial effect on the predicted dynamic behavior of a reactor.
8007240 0go (y
' 3.
Has the Staff ever determined the " difference" mentioned in Interrogatory 2 above for any BWR core used as a production facility?
Response
No.
4.
Has diffusion theory calculation ever been used to determine the outcome of any of the design based power excursion accidents for the ACNGS?
Response
Yek 5.
WASH-ll46, pp. III-94, states, "The experimental basis for determining the physical correctness of space-time calculational results (for large cores) does not exist as it does for small cores." As this publication is several years old, does staff believe there have been progressive steps which would modify this statement today? If so, what progress has been made?
Response
The exact context in which the statement cited in this interrogatory is made is not known. However, some coments can be made:
(1) If the concern is the rod drop accident, it should be pointed out that this is not a large core i
event even if it takes place in a large core. The primary excursion is a purely local phenomenon occurring in the vicinity of the dropped rod.
(2)For full core events some recent experiments have been made.
In particular the Peach Bottom tests 3 ncluded axial and radial power history studies.
i These 3
EPRI-NP-564
" Transient and Stability Tests at Peach Bottom Atomic Power Station, Unit 2, at End of Cycle 2," June,1978.
' studies and others, such as xenon transient and thennal hydraulic stability studies serve as benchmarks to compare with the r ;sults of dynamics codes.
6.
Does Staff believe the WIGLE code accurately predicts the neutron wave characteristics of any of the ACNGS design based power excursion accidents?
Response
The WIGLE code is not used in any analysis for any transients for the ACNGS reactor. Therefore its suitability is irrelevant.
CONTENTION 11 - Sper.t Fuel Pool 1.
Does Staf f rely on the Sandia mathematical model SFUEL in calculating spent fuel heatup in a loss of water accident (LOWA)?
Response
A.
No. The Staff.does not postulate a total loss of water in the spent fuel pool. We verify that sufficient defense-in-depth is provided in
(
the spent fuel pool design to assure an adequate water level and proper spent fuel cooling at all times.
For Allens Creek this includes:
1.
A seismic Category I concrete pool.
2.
A stainless steel pool liner.
3.
A safety grade spent fuel pool cooling system.
4.
Backup pool cooling provided by the safety grade RHR system.
5.
A safety grade makeup water supply to the pool.
6.
A piping arrangement with connections to the pool which prevents draining the pool below a safe level.
b
\\
4-7.
A leak detection system and redundant pool water level and temperature instrumentation.
B.
The Allens Creek PSAR and SER were used as reference in preparing this response.
C.
N/A.
D.
Standard Review Plan (SRP) Section 9.1.2 and 9.1.3.
2.
Does Staff take the position that if there is loss of water, the zirconium fuel cladding would not ignite from the spent fuel pellets when the rods become uncovered?
Response
Yes, experiments (Ref.1-5) have demonstrated that massive zirconium and Zircaloy do not ignite in steam and do so in high pressure pure oxygen only if there is no oxide film present on the surface of the metal, regardless of the temperature of the metal. Since the oxide is soluble in the molten metal, the molten metal can ignite in oxygen if the oxide can be kept from covering the surface.
But even molten zirconium will not ignite in steam, as the heat of reaction is not high enough to overcome the heat losses to the surroundings.
1.
" Zirconium Fire and Explosion Hazard Evaluation," USAEC Report, TID-5365 August 7, 1956, 2.
F. E. Littman, F. M. Church, and E. M. Kindennan, "A Study of Metal Ignitions II. The Spontaneous Ignition of Zirconium," Journal of Less-Common Metals, 3_, p. 379-397 (1961).
3.
L. F. Epstein, " Correlation and Prediction of Explosive Metal-Water Reaction Temperatures," Nuclear Science and Engineering,10, p. 247-253 (1961).
j af
4 6 4.
L. Baker and L. C. Just " Studies and Metal-Water Reactions at High Temperatures III. Experimental and Theoretical Studies of the
?
Zirconium-Water Reaction," Argonne National Laboratory Report, ANL-6548, May 1962.
5.
N. I. Sax, " Dangerous Properties of Industrial Materials," Van Nostrand Reinhold Company, 1975.
3.
Regardless of the Staff's conclusion in part 2, does it believe that no heat-up will occur in LOWA conditions to pemit melting of fuel? If no, please indicate why.
Response
Yes. The spent fuel pool is built to seismic qualifications that are requiret.
of Class I structures, and the Staff does not consider a LOWA to be a credible accident.
Fuel melting, however, could occur if (a) water was lost from the pool or (b) heat was generated beyond the capability of the SFSP cooling system.
Of these two events, the fomer is not considered credible, and the latter would not result in fuel damage until the pool cooling water could be boiled off.
The time required for boil off is about 3.8 weeks for an average pool inventory of 50,000 cubic feet at a loading of 1.6 Mw (Ref.1). Obviously, the slow nature of this f. vent will pemit makeup water to be obtained from several sources, even from offsite.
1.
U. S. Nuclear Regulatory Conunission, " Reactor Safety Study: An Assessment of Accident Risks in U. S. Commercial Nuclear Power Plants,"
USNRC Report NUREG-75/014 (WASH-1400), Appendix I, Section 5.1, " Spent Fuel Storage Pool (SFSP)," October 1975.
4.
If the answer to (#3) above is yes, would not a fuel mass containing plutonium, not be capable of excursion which would produce a strong nuclear explosion?
h
' Response No, the fissile concent~ ations or uranium and plutonium in spent commercial r
fuel are insufficient to pennit " strong nuclear explosions."
5.
Has Staff ever analyzed criticality accidents resulting from missing " neutron absorbing materials"?
Response
The effect of missing some of the neutron absorbing material (usually boron carbide) between storme locations in the pool is sometimes investigated by applicants. However, the main thrust of NRC efforts in this area is in the direction of assuring that the material is there.
In situ tests are required (or other quality assurance procedures) to assure that no more than a specified number (usually one) absorber plates is missing to within some high probability at high confidence level (usually 95/95%).
6.
What is Staff's believe or findings for when heated zirconium is re-covered with water, in an attempt to refill the spent fuel pool (SFP) following partial or full loss of water contents?
Response
In general, the rate of heat removal from the fuel rod would be increased and the fuel rod temperature would be decreased. We do not understand whether there was an intent to request a description of a specific scenario.
i P
. 7.
Does Staff concur currently with the WASH-1400 conclusion that i
only new spent. fuel will release radioactivity in the event of melting in and LOWA? If yes, are there any studies other than assumptions on which Staff relies?
Response
WASH-1400 states (1) that the decay heat levels in freshly unloaded fuel assemblies may be sufficiently high to cause fuel melting if the cooling water is completely drained from the spent fuel storage pool (SFSP). We agree with this statement. The report further stated that the decay heat levels would decrease during storage time in the pool.
It was estimated that, on the average, fuel in the pool will have undergone about 125 days of
. decay. Although the report questions the likelihood of such fuel melting, it did not limit such a possibility to oni', freshly unloaded fuel assemblies.
Rather, to assure that the risk wot
..ot be underestimated, the study assumed that the decay heat levels in even older s ent fuel assemblies may be suf-e ficiently high to cause fuel melting if the cooling water were completely drained from the SFSP. We are not aware of any studies which indicate that this assumption is non-conservative.
(
l.
U. S. Nuclear Regulatory Commission, " Reactor Safety Study: An
/
Assessment of Accident Risks in U. S. Comercial Nuclear Power Plants,"
l USNRC Report NUREG-75/014 (WASH-1400), Main Report, Section 3.5,
" Accidents Involving the Spent Fuel Storage Pool," October 1975.
t 8.
What is the Staff's criteria for a "high density" as opposed to other density SFP storage rack?
l
' _ Response A.
The Staff's criteria for a "high density" SFP storage rack are the same as that for other density SFP storage racks.
B.
N/A.
C.
N/A.
D.
N/A.
9.
Will Applicant be required to remove the fuel channels surrounding each fuel assembly prior to storage in the SFP, as expected in NUREG/CR-0649 (SAND 77-1371)?
Response
A.
No. The Applicant is not required to remove the fuel channels on spent fuel assemblies prior to their being placed in storage. While it is not required, it is a routine practice to remove the channels from spent
'uel assemblies.
B.
TJe Allens Creek PSAR was used as reference in preparing this response.
C.
N/n.
D.
SRP Section 9.l.4.
- 10. Has Staff undertaken to study or cause to be studied the consequences of a spent fuel pool LOWA in a BWR, as opposed to the consequences for all nuclear power plants?
l
Response
i A.
No. The Staff has not undertaken nor caused to be undertaken arty studies to assess the consequences of a spent fuel pool " loss of water accident" for a BWR, other than that information included in NUREG/CR-0649.
p
- B.
N/A.
3 C.
N/A.
D.
N/A.
CONTENTION 12 - Rod Control and Information System (RCIS) a 1.
Up to what percent of power will Applicant be required to use the RCIS?
Response
The RCIS system will be required to be in use from cold startup conditions to full licensed puwer. However, it will be in the Rod Pattern Control System made from cold startup ta approximately 20 percent of full power. Above 20 percent of full power the RCIS will operate in the Rod Withdrawal Limiter Mode.
2.
Does Staff currently believe all substantial problems with the previous Rod Manual Control System have been eliminated by the addition of a second rod position information system and redundant rod action systems as in RCIS?
Resporse The Staff believes that the addition of redundancy to the RPCS portion of the RCIS, coupled with the adoption of the Banked Position Withdrawal Sequence, and in view of the very low probability of a Rod Drop Accident, has reduced the: potential for adverse health and safety effects to the point that further improvements are not warranted.
8 3.
As a result of the Three Mile Island investigations and studies has Staff detennined if any rules or modifications to RCIS systems are in order? Particularly, has there been any effort to attack the problem of operator initiated bypass due to spurious APRM signals of surplus neutron flux which reportedly arise fron " uranium dust."
Response
The Staff's studies to identify any additional rules or modifications for ACNGS as a result of its Three Mile Island investigations have not been completed. When completed, the results will be reported to the Applicant with ccpies to the parties in the hearing proceeding.
5.
If the ACNGS is operating at 10% power and critical, is the maximum worth of any insequence control rod which is not electrically disanned less than 0.10 delta k?
Response
Yes.
6.
When the reactor is above 10% design power what is the maximum worth of any control rod, including allowance for a single operator error?
Response
An overestimate of the maximum worth of a rod with a single error
'n be ebtained from Figure 3-11 of Supplement 1 of NED010527. The numbar from
- that figure is 0.011 ok.
_11 -
' CONTENTION 15 - WIGLE code i
1.
Does Staff agree that the Adiabatic Prompt Excursion Model featured in NE0010,527 is "relatively accurate" (page 4-1)?
If so, relatively accurate to what other models?
Response
The Staff believes that the Adiatatic Prompt Excursion Model is conservative with respect to models which account for prompt gamma ray heating of the moderator. (See BNL-NUREG 27544 for a discussion of the effect of prompt heating). Further, the particular model described in NED0-10527 assumes a constant axial power shape during the excursion--a further source of con-servatism as compared to two dimensional models such as that in the BNL-TWIGL code.
(Page 6) 1.
Please cite a scientific paper or report which unequivocally states that space-time neutron kinetics theory is adequate for describing and predicting the neutron behavior exhibited in a reactor core more than 10 feet in width.
Response
Two reports may oe cited which show good agreement between calculation and experiment for a transient in 15 ft. diameter reactor (Peach Bottc= 2 cited in response to Interrogatory 5 on first page of these interrogatories).
These reports are:
l'.
NED0-24154, Vols. 1, 2, 33 " Qualification of One-Dimensional Core Transient Model for BWRs".
3 ol. 3 is Proprietary.
V
~~
(
- 2.
BNL-NUREG-26684
" Analysis of Licensing Basis Transients for a BWR/4" j
Sept. 1979.
2.
Referring to (1) above, you may cite one which describes for any size core, if there is none for (1).
If there is, omit this.
Response
See response to Interrogatory 1 (page 6).
3.
What was the outcome of NRC review of General Electric program to analyze the control rod drop accident and revised NED0-10,527 into a three dimensional code?
Response
So far as we are aware GE has no plans. to replace the Method of NED0-10527 with a three-dimensional calculation (it would not be possible to revise the code to be three dimensional). The Staff has not required any vendor to resort to three-dimensional studies for reactivity insertion accidents. We have been concerned at times with the effect of dimensional order on such calculations and are pursuing this on a low priority basis. The reasons for our lack of great concern are given in a memorandum (Rusche, NRR to Fraley, ACRS, date June 1,1976) on ACRS Generic Item IIA-2, Control Rod Drop Accident.
~
4.
Does the General Electric code description (4-11 of SER) con-sider outlying region (more than 1 migration path) from a local perturbation?
. a
Response
The method described in NED0-10527 made use of " outlying regions" to determine the rod worths used in the calculation and to determine the Doppler weighting factor for the calculation. The dynamics calculation itself is of course a point-kinetics calculation.
5.
In the NRC summary of NED0 20,953A, the flux distributions are said to be primarily determined by high energy neutrons.
Would these be all neutrons whose energy is greater than 1 Mev in this group and no others?
Response
In the context of nodal calculation code such as described in NED0-20953A the term fast neutrons includes all neutrons above the resonance region, i.e.
above about 10 kilovolts in energy.
6.
In an effort to understand nomenclature, would it be a fair and correct description of the code which is the subject of Staff's reply to Interrogatory #5 of this Intervenor's 4th set of Interrogatories to call it a one dimensional neutron kinetics model with reactivity feedback from con-trol rod, voids and Doppler effects, and that it ignores Dancoff factors?
Response
Yes, except Dancoff factors are treated in the treatment of Doppler effects.
7.
Can delayed neutron contribution be neglected in WIGLE reactor excursion calculations?
r
Response
Generally, no.
If the excess reactivity inserted is sufficient to exceed prompt critical by a large amount, then it may be ok. Kinetics calcula^ ions are not generally very sensitive to the delayed neutron fraction, however, 8.
Does Staff believe approximations to transport theory are usually quite satisfactory in calculating power excursion neutronic effects?
Response
Yes.
CONTENTION 25 2.
Does Staff believe at this time that in order for the zircaloy clad to melt during a power coolant mismatch (PCM) such as a flow blockage event, the melted fuel must be ex-truded sufficiently to make contact with the clad? If other intra clad events during a PCM are considered possible initiators of clad melt, please mention them in your answer.
Response
In answer to the question of whether molten fuel must contact the cladding as a requisite for cladding itself to melt during a postulated power coolant mismatch, the cladding would, of course, melt whenever its temperature reached the melting point of the Zircaloy, by whatever mechanism. Since the melting point of Zircaloy ( 1850 C) is much lower than that of UO2 ( 2800 F), and since the thermal conductivity of U0 is very low, it is theoretically possible 2
to (a) melt the cladding before the U02 pellets, or (b) have some centerline melting of U0 while solid UO contacts the cladding (which may or may not be 2
2 w
molten). TW, if the coolant flow were throttled back sufficiently (on the order of 90% or more) so as to permit steam blanketing of the cladding while at the same time rod power remained sufficiently high, cladding could melt while the UO2 pellets remained essentially solid. However, the fuel assembly design and core operating and safety features are such as to make such a scenario very unlikely.
s 3.
Can you supply a reference which describes an incident (probably in 1973) written-up in Nuclear Safety 15(1),
- p. 37, where "a small piece was missing from the corner of one channel in the Swiss BWR) KKM reactor.
If not, any additional infonnation with regard to the potential for such pieces to reach the fuel bundle inlet or to remain in the fuel bundle as estimated at that time would be appreciated.
R_gsponse We have no information regarding a channel box problem in a Swiss BWR other than what was presented in Section 2.1.1 of the " Safety Evaluation by the Directorate of Licensing, U.S.A.E.C., Relating to Channel Box Wear in the Vermont Yankee Nuclear Power Station (Docket 50-271) and the Pilgrim Nuclear Power Station (Docket 50-293)," October 26, 1973.
4.
In October,1973, cracks estimated 18 in. long by 0.25 in, wide were discovered in channels in the Vermont Yankee BWR.
Were all of these decided to be splits in the material or were there pieces believed to have vacated these channel sheets, such that they would join the coolant flow? In other words, what were the results of investigations. What steps were taken to prevent the further occurence of these defects?
I
. Response A safety evaluation report regarding the BWR channel box wear problem was issued in October 1973.
(See response to Q.4.) As indicated therein, no pieces of channel wall were missing in any of the channels inspected. Sub-sequent to that report the Vermont Yankee Nuclear Power Corporation submitted a report entitled " Summary Report on Vennont Yankee Channel Wear Investigation and Corrective Measures Taken." That report describes the action taken to prevent recurrence of the damage. Because the cause of the channel wear was identified to be the interaction of high velocity flow from the flow bypass holes with the temperatory control curtains, (an interaction that caused the curtains to vibrate and damage adjacent channels), the solution was to eliminate the high velocity flow by plugging the bypass holes.
CONTENTION 27 1.
What was the temperature of the pedestal area during the Three Mile Island Accident, which started March 29, 19797
Response
The temperature of the pedestal was not measured. The temperature of the reactor building atmosphere reached 170 F (page 1A.29 of NUREG-0600, "Investi-i gation Into the March 28, 1979 Three Mile Island Accident by Office of In-spection and Enforcement").
2.
Has General Electric provided a means by which the temperature reached by the pedestal following a design based LOCA is to be calculated, such that damage can be detennined?
_ 17
Response
Not for ACNGS.
3.
Will Applicant be permitted to leave the space between the concentric steel rings which support the reactor vessel empty, as Applicant implies it can in 3.8.3.1.7 of of the PSAR (Am. 35)?
If so, must the space be sealed to prevent corrosion?
Response
This question is not applicable since the Applicant's proposed design includes use of concrete for filling the space between the concentric steel rings (note that the use of the word "may" in Section 3.8.3.1.7 appears to be an editorial error since the remainder of the PSAR text and figures state and show that concrete will be used.
4.
Are any other filler materials used in RPV pedestals than concrete?
Response
No.
CONTENTION 38(b) - Cold Shutdown in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.
Would branch Technical Position APCSB9-2 " Residual Decay Energy for LWRs for long term cooling" have to be modified in its representative operating history of 46,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br />; if higher burn-up fuel is placed in the core of the reactor?
. Response i
Not for the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time period.
CONTENTION 39 - Fuel Rod Ballooning 1.
Of course Applicant's NSSS uses lower pressure than BWR systems, but does Staff believe the pressure within a fuel rod must reach an absolute amount before an excess of pressure (that is hoop stress) will being to cause fuel rod swelling?
Response
Fuel rod swelling (i.e., elastic plus plastic circumferential straining) begins at the instant when the fuel rod gas pressure exceeds coolant system pressure. The rate of swelling is complexly dependent on several interactive variables such as the following:
1.
The amount 4 differential pressure (or cladding hoop stress).
2.
Ciadding temperature.
3.
The presence of embrittling agents (hydrogen, oxygen, iodine, etc.)
in the cladding.
4.
The rate at which the differential pressure is changing (cladding strain rate).
5.
The presence of crack initiation sites (incipient defects, oxide cracks).
2.
Does Staff believe that for events such as ATWS, the rapidity of the power ramp would lead only to brittle fracture and not plastic strain of the fuel rod cladding? If so, please cite a reference supporting this that you believe would be likely available easily.
. Response h
Fuel vendor generic topical reports that describe the ATWS events report that the degree of calculated cladding plastic strain will be less than 1%. This issue is presently under review, but the Staff has not found any reason to reject vendor conclusions on the degree of strain. At present, there are no non-proprietary versions of these reports.
3.
What is the role of fuel burn-up in relation to strain or plastic defomation on a fuel rod typically?
Response
For fresh fuel, the cladding will strain inward (creepdown) due to the coolant system pressure exceeding the fuel rod pressure. Then with time, the fuel rod pressure will increase due to the creepdown and fission gas release to the gap inventory. Also with burnup, fuel pellet swelling occurs and may possibly press against the cladding wall and strain the cladding outward.
For consideration of LOCA and ATWS event, higher burnup fuels will have larger concentrations of embrittling agents, which tand to limit ballon 'ng s trains.
4.
Is strain (or defomation) linearly related to hoop stress or is there variation in the amount of strain for particular hoop stress pressure ranges?
Response
The relationship of strain to cladding hoop stress is not particularly linear over any significant stress interval.
20 -
5.
How much clad swelling (% of outside diameter of an ACNGS fuel rod) can occur without significantly impairing the effectiveness'of the ECCS, using 10 C.F.R. as guidance for the tenn effectiveness?
Response
As set forth in 10 C.F.R. 50.46, the ECCS is required so that in the event of a LOCA the following criteria will be met:
1.
Maintain peak cladding temperature below 220"F.
2.
Limit maximum cladding oxidation to 17%.
3.
Limit core hydrogen generation to 1%.
4.
Maintain a coolable geometry.
5.
Provide for long-term coolirg.
Any positive increment in calculated cladding strain that results in a reduction of the previously predicted margin to any of the above listed criteria is an impairment on the effectiveness of the ECCS, even though the criteria may still be met and the LOCA analysis entirely acceptable to NRC.
Specifically, cladding strain influences the effectiveness of the ECCS by the following mechanisms:
1.
Increasing the cladding heat transfer surface area which enhances convection heat transfer.
2.
Decreasing the fuel pellet-to-cladding gap conductance which reduces the heat transferred out of the fuel.
i
3.
Increasing the cladding surface area which increases the oxidation rate and increases the rate of cladding heating.
4.
Reducing the coolant flow area which degrades the convection heat transfer.
5.
Increasing or decreasing the rod-to-rod radiation heat transfer.
The first two mechanisms enhance, the second two mechanisms impair, and the last mechanism i.e enhance or impair the effectiveness of the ECCS.
Hence, without running the ECCS evaluation code with the plant-specific details, it is not possible to conclusively detennine a priori the exact impact of cladding strain on the effectiveness of ther CCS.
6.
What administrative procedure would be followed if mid-Construction of ACNGS, a large number of the Three Mile Island -II Fuel rods were found to have been blocked by swelling during the 1979 incident?
Response
The NRC believes (Refs.1 and 2) that a large number of the TMI-2 fuel rods ballooned and ruptured and that a large percentage of the core was blocked during the March 1979 accident.
It is expected that the fuel examination of the TMI-2 core will provide data that will subsequently be useful to NRC in verifying the degree of cladding swelling that is anticipated for small-break LOCAs.
Fuel behavior information gained from the TMI-2 fuel examination that might indicate a deficiency in the ACNGS PSAR will be available prior to the ACNGS power operation and probably prior to the time that the ACNGS fuel is i
o l
- manufactured, thus allowing sufficient time for operational or manufactural j
alterations that might be necessary for the ACNGS to operate in compliance with existing regulatory requirements.
1.
U. S. Nuclear Regulatory Commission, " Evaluation of Long-Term Post-Accident Core Cooling of Three Mile Island Unit 2," USNRC Staff Report NUREG-0557, Appendix A, " Core Damage Assessment for TMI-2 "
May 1979.
2.
Nuclear Regulatory Commission Special Inquiry Group, "Three Mile Island: A Report to the Commissioners and to the Public," Volume 2, Part 2, Chapter C, " Plant Behavior and Core Damage," January 1980.
7.
Are the effects of partial depressurization of the RPV such as from a stuck open SRV believed likely initiators of swelling in BWR rods based on a mechanism of FR pressure exceeding the RPV pressure?
Response
If a SRV was stuck open and there was no makeup charged to the primary coolant system, then partial depressurization would occur and such could be considered as a small-break LCCA.
Fuel rod swelling could then result after the rod internal pressure exceeded coolant system pressure.
8.
Does Staff accept the NUREG/CR 0269 (" LWR Fuel Response During RIA Experiments, 3/78) Sec. 3.2.1 and See Figure 27 page 50 statement for irradiated water logged PWR rods, that clad ballooning is observed at low energy depositions of 50 to 70 cal / gram? Does Staff accept this for BWR rods?
?
Response
j Yes, such failure at low energy depositions is believed to be due to higher internal rod pressures (i.e., steam is generated in water-logged rods and fission gas is released to the gap inventory in pre-irradiated rods). Yes,
.. the present the Staff has no reason to reject the relevance of these RIA experiments to BWR cores.
CONTENTION 24 - Rod Drop Accident 1.
In the control rod drop accident, what basis or scientific report or reports show 100% of noble gases and 50% of iodines (See Page 15-17 of NUREG 0152) would escape the fuel rods, but no fuel fragments whatsoever?
_ Response The structure of'the question "in the control rod drop accident, what basis or scientific report or reports show 100% of noble gases and 50% of iodines...
would escape the fuel rods, but no fuel fragments whatsoever" indicates that the criteria required for the assessment of the radiological consequences of BWR control rod drop have been misinterpreted. There is no scientific " basis" or reason to believe that 100% of the noble gases and 50% of the iodines would escape the fuel rods. That is simply a number that is believed to be bounding and which is required for licensing calculations.
In fact, as indicated in Acceptance Criterion le of Section 15.4.9 (Appendix) of the Standard Review Plan, that assumption was originally intended to apply to only that fraction of the fuel which would be molten, whereas in current NRC practice the assumption
- is applied to all fuel rods that are calculated (or assumed) to fail, even
)
if the failure is no mo're than a pin-hole.
Using current criteria (for fuel enthalpy) no rods are predicted to fail, but GE assumes for conservatism that 770 rods will fail. Thus, for this analysis, conservatisms are compounded along the way to provide a radiological dose estimate that is believed to be significantly higher than would actually occur.
With regard to fuel fragmentation the fuel peak enthalpy limit is intended to insure coolable geometry (i.e. preclude fuel *ragmentation). Also, see the response to Interrogatory 8, Contention 46.
CONTENTION 29 - Ultimate Heat Sink Inadequacy 1.
Does Staff expect to do final design analysis of Applicant's Ultimate Heat Sink prior to the construction permit hearings?
Response
No.
2.
In a letter of 6/23/78 from F. R. Brown, Engineer, Corps of Engineers, Waterways Experiment Stations, Vicksberg, Miss.,
l to Gammill (NRC, Site Analysis Branch) he states that relative to compaction requirements for Class I-a Fill, an acceptable criterion for ACNGS is either 80% relative density or 95% modified Proctor density whichever is greater should be provided to prevent liquifying of this material from a safe-shutdown earthquake. Has Staff eliminated any uncertainty about meeting this or has it decided it cannot follow this admonition? (Letter is in theACNGSdocket.)
/.
Response
Yes. This matter was 1.isted as an unresolved issue in Supplement No.1 to the Safety Evaluation Report (Item II.7 of Section 1.1 and Section 2.5.4.2).
The solution was reported in Supplement No. 2 to the Safety Evaluation Report (page B-4 of Appendix B and Section 2.5.4(1)).
CONTENTION 45 - Later core support 1.
In NUREG/CR-1018, p.13 (D-3(c)), the contractor report states that a later LOCA force requires an additional margin of support in the fuel assemblies above that for the SSE by about 30%.
This lateral force is due to,
" flashing which occurs near the end of the sub-cooled blowdown portion of the LOCA transient," and the report suggests it should be incleded in the LOCA analysis.
(a) Does the ACNGS core contain features providing this margin?
(b) Has the NRC developed any rulemaking plans to con-sider this lateral force?
(c) Would the insertion of concrete (see: Doherty Contention #27) into the RPV pedestal (as a possible plan of Applicant) give additional later support, or is the weakness regarded to be between core internal parts?
(d) Does this same flashing occur in the ACNGS core or actuation of the ECCS following a LOCA?
Responst Or'iginal Contention 45 uses NUREG/CR-1018, " Review of LWR Fuel System Mechanical Reponse With Recommendations for Component Acceptance Criteria,"
by R. L. Grubb of EG&G Idaho as a reference. The page cited recommends that a factor of 1.3 be applied to lateral LOCA (Loss-of-Coolant Accident) loads eh the fuel due to steam flashing.
~
.e.
- It is unfortunate that the limits on the application of the recommendation were not discussed in the recommendation. However, the previous paragraph does refer to Appendix B of the report, which is a paper entitled " Analysis of Three-Dimensional Effects on PWR Blowdown Heat Transfer," authored by P. North, R. L. Benedetti, and L. V. Lords. This paper states the assumptions used in the analysis and these assumptions limit the applicability of the results. Based on these assumptions, and a knowledge of the fundamental differences between BWRs and PWRs, the flashing loads safety factor does not apply to BWRs for the (411owing reasons:
First, as Appendix B states, the system modeled was a PWR with a double area cold leg pipe break. A PWR operates below the boiling point (subcooled) where a BWR operates at the boiling point of the water coolant. As stated in Contention 45, the lateral (fuel) force is due to " flashing which occurs near the end of the subcooled blowdown portion of the LOCA transient." Since the BWR does not operate in the subcooled regime like the PWR analyzed, it is not possible to have these flashing loads in a BWR.
CONTENTION 46 - Xenon Transients 1.
Does Staff agree with this statement, taken from Page t.3-29 of the Montague PSAR, "BWRs do not have instability problems due to xenon?
4
. Response Yes.
2.
In a Licensee Event Report (LER) on the Brunswick, Unit 2, the utility stated as a cause for the event, "The inner filter on the drive is believed to be loose and engaging the uncoupling rod, uncoupling the blade from the drive unit." How can a filter uncouple these parts?
I Have never been fortunate enough to see any drawing showing an uncoupling rod in ACNGS or another BWR. Can you provide a Reference?
Response
An unpublished report, " Interim Report - Dresden 2 Control Rod Drive Performance Problems," November,1977 is enclosed to provide you with descriptive information.
3.
According to Tennessee Valley Authority's response (7909280041) to Item 1 of I.&E. Bulletin, 79-12, there is an uncertainty of i 0.3% in reactivity in estimating rod worths when its Browns Ferry plants are at 10% or less power.
Is this true for ACNGS also? The statement is evidently backed by Boston Edison's response for Pilgrim I (7909110005).
Response
The 0.3% in reactivity cited by TVA in their response is the uncertainty in predicting the critical rod configuration for startup. This number is also cited in Topical Report NED0-2C946A for cold startups in BWRs. This would be expected to be the number for ACNGS.
v 4.
When d.cre is a xenon transient at the bottom of the core would the source range monitor SRM, located midway from top and bottom present a middle estimate of the in-duced neutrons from a rod fall of a single or two notches?
- Response Xenon transients are whole core phenomena and so a transient occurring just in the bottom of the core is not possible. However, the magnitude of the changes in xenon concentration may vary from point to point in the core.
The effect of the xenon concentration in the vicinity of a source range monitor would be negligible on the rate of change of the response but not negligible on the amplitude of the change. An SRM located at the core midplane would be expected to respond to the core average change in neutron level for a symmetrical axially varying xenon concentration--not otherwise.
5.
Would use of the Traversing Incore Probe be specified by current operating instruction to prevent this today?
Response
The use of the Traversing Incore Probe during the startup range of power levels is not specified. The sensitivity of the TIP's is too small to make their use reliable in this range.
6.
According to a letter from Elbert Eppler, (7910020637), a nuclear consultant, of 7/9/79, the " unlatching" between control rod and control rod drive occurs at notch 48, the last notch going toward complete removal.
I believe he meant most often this was when such uncoupling occurred.
Has Staff any explanation for this?
Response
See pages 2-10 and 2-11 of the reference provided in response to Interrogatory 46-2.
'. 7.
Eppler's letter (see "6" above) refers on Page 7 to
" bypassed voids" as a new problem in rod withdrawal in a BWR despite group withdrawal of rods as abdicated by NEDO 21,231. What are "by-passed-voids"?
Response
Eppler's letter should probably refer to bypassed rods. The effect of by-passed rods on the potential dropped rod worths in the Banked Position Withdrawal Sequence has been treated in NED0 21231. Technical Specifications permit a total of eight bypassed rods and imposes certain res trictions on their location. NED0 21231 shows that if the latter restrictions are ignored and the light rods to be bypassed are all in a single quadrant of the core (thus maximizing their effect) the potential dropped rod worths are still below that required to produce 280 cal / gam.
8.
Does Staff currently support the position of NED0 20, 948:
"There are no experimental data to date that indicate a possibility of prompt fuel failure in the fuel enthalpy range discussed. Thereforr the peak fuel enthalpy and design limit of 280 cal /gr: is considered justifiable and conservative"?
Response
The 280 cal /9 radially aver.ged peak fuel enthalpy acceptance criterion is intended to serve two purposes:
1.
To prevent pressure pulses that might result from prompt fuel element rupture and threaten the primary system boundary, where " prompt rupture" is defined as "a rapid increase in internal fuel rod pressure
. due to extensive fuel melting, followed by expulsion of molten fuel and dispersal of fuel cladding fragments into the coolant" (Reg.
Guide 1.77).
2.
To prevent the loss of "coolable geometry," or coolability.
With regard to the first of the two above-stated purposes for the 280 cel/g criterion, we believe that there is sufficient experimental evidence from the SPERT and PBF tests to provide adequate assurance that " prompt rupture" will not occer as a result from a BWR control rod drop accident that produces an energy depos? tion (280 cal /g radially averaged fuel enthalpy. Thus pressure pulse damage to the crimary system is unlikely at such energy levels.
With regard to coolable geometry, we are not fully confident that the 280 cal /g limit will ensure coolable geometry, i.e., that rod-like geometry will be retained. Preliminary results from PBF indicate that unirradiated and irradiated rods may have fragmented at an energy (peak enthalpy) below the 280 cal /g level.
In PJdition, it WPS reCently discovered that the early SPERT test results were reported as total energy (integral of reactivity pulse) rather than peak fuel enthalpy.
For SPERT, 280 cal /g total energy corresponds to about 230 cal /g peak (i.e., maxima radially averaged) fuel enthalpy.
It is thus tempting to recommend an interim criterion of 230 cal /g peak (radially averaged) fuel enthalpy. However, while this value would pre-clude incipient melting in RIAs originating from zero power, it would t
e
not pre'clude melting in full-power RIAs. A preferable interim criterion might be to avoid incipient local melting; that is, a new lin.it that might be considered could be 267 cal /9 local peak fuel enthalpy, since it would indeed preclude local melting (which is what is really intended).
Regardless of whether 230 cal /g radially averaged peak fuel enthalpy or 267 cal /g local peak fuel enthalpy is used, recent BNL analyses of the BWR rod drop that for the first time take into account the effects of moderator thermal / hydraulic feedback, indicate consequences much lower than standard (GE) non-feedback (hydraulic) methods. Hence, neither the 280 cal /g, nor any new limit currently being considered, would be exceeded.
CONTENTION 33 - Inadequate estimate of Doppler effect 1.
Is G. E.'s treatment of Doppler effect in full compliance with Appendix A. Part 8 of Reg. Guide 1.77., " Assumptions used for Evaluating a Control Rod Ejection Accident for a PWR7 (a)
If not, what differences are there than those strictly due to the fact ACNGS is a BWR7
Response
G. E.'s treatment of the Doppler Effect, as described in NED0-10527 meets the criteria of Reg. Guide 1.77 for this effect.
f
2.
Doppler broadening is said to produce a negative temperature coefficient, (that is reduce reactivity with increased temperature).. However, the removal of reactivity is said to not be due to a heat transfer process. What is the theory or experimental data that explains how temperature of the neutrons can be changed without a heat transfer process? Note to critics:
I seek to confirm suspicions on this!)
Response
Heat due to fission is produced i, the atoms of U-235 - U-238 and plutonium since these are intimately mixed in the fuel no heat transfer mechanism needs to be invoked in order to get heat into the U-238. The broadening of the resonances is due to heat-induced motion of the U-238 atoms and not due to heat transferred to the neutrons.
3.
On page 2 of NEDO 20-964-1 (a G. E. report) it states,
"[I]ndependent of overlap, small variations in Eippler reactivity can occur due to differences in the con-centrations of nuclides which compete with U-238 for resonance neutrons. These effects are reasonably well represented in Doppler computations."
(a) Has Sta"f reviewed this report and agreed with this co1clusion fully?
(b) What has Staff asked G. E. to do to better represent the nuclides said to compete with U-238 for resonance neutrons?
(c) What is the name of code used by G. E. to compute the Doppler reactivity variations? Please cite literature on its adequacy.
L
Response
The presence of nuclides other than U-238 in the fuel has an effect on the Doppler reactivity. A case in point is the presence of Pu-240 which changes temperature with the U-238 and contributes to the Doppler coefficient, making it more negative as the fuel burns up. Other fission products may have the effect of reducing the coefficient or of increasing it. The interactions among the various items are very complex but in any case the effects 'are small.
They are accounted for by a multiplicatics factor'(generally 0.95) on the calculated Doppler coefficient before using it in safety evaluations.
The name of the code used to calculate Doppler coefficient variations is not known.
It is, however, described in NED0-20964 and NEDE-20913-P (Proprietary).
GENERAL INTERROGATORIES 1.
Are any of the so-called "Michelson Concerns," raised by an engineer from TVA about and in the wake of the TMI-2 event considered relevant to BWRs by Staff? If so, please list the concerns in phrases sufficient to identify if they are relevant to any contentions in this proceeding raised by Intervenors.
Response
New and revised requirements of the Nuclear Regulatory Commission resulting from TMI-2 event investigations have not been stated for construction permit applications. When issued, these requirements will be forwarded to the ACNGS Applicant with copies to all parties.
e
' CONTENTION 35 - Welder Adequacy g
1.
Referring to I.&E. Report 50-498/79-08 (South Texas Nuclear Project) is this a common item of non-compliance with projects of this sort?
Response
Yes. Violations characterized as a failure to have prescribed documented instructions, procedures or drawings, of a type appropriate to the circum-stance, for activities affecting quality in accordance with Criterion V of 10 C.F.R. Part 50, Appendix B are common among Notices of Violations issued by the NRC.
CONTENTION 47 - Turbine Missiles 1.
Does Staff believe the current level of failure by throwing of turbine blocks is so low that no modifications need be done as to orientation or structure of the ACNGS power block?
Response
As stated in Section 3.5.3 of Supplement No. 2 to the Safety Evaluation Report, the Applicant has changed the orientation to a peninsular orientation of the turbine generator relative to the containment, auxiliary and control buildings.
The Staff did not require a peninsular arrangement relative to the radwaste building for the reasons stated there.
/
' 2.
Has NRC considered requiring the availability of spare spindle after a certain period of time in anticipation of cracking or other turbine defect?-
Response
4 No.
t 1
L 4
1 4
l T
g,
...=,- _ __ - -
,,,r,-
,.. - - -. -, - -, -. ~
C UNITED STATES =?F AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD k
In the Matter of HOUSTON LIGHTING & POWER COMPANY J
Docket No. 50-466
)
(Allens Creek Nuclear Generating
)
Station, Unit 1)
)
AFFIDAVIT OF CALVIN W. MOON I hereby depose and say under oath that the foregoing NRC Staff responses to interrogatories propounded by John F. Doherty were prepared by me or under my supervision.
I certify that the answers given are true and correct to the best of my knowledge, information and belief.
0 W,%w Calvin W. Moon ~
Subscribed and sworn to before me this 21 t day of July, 1980.
s 2
LL ctaryfublic
/
My Commission expires: July 1, 1982
UNITED STATES OF AMERICA
[
NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 5
In the Matter of
)
)
HOUSTON LIGHTING & POWER COMPANY Docket No. 50-466 (Allens Creek Nuclear Generating
)
Station, Unit 1)
)
CERTIFICATE OF SERVICE I hereby certify that' copies of "NRC STAFF'S PARTIAL RESPONSE TO JOHN F. DOHERTY'S THIRTEENTH SET OF INTERR0GATORIES" and " AFFIDAVIT OF CALVIN W. M00N" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or, as indicated by an asterisk, through deposit in the Nuclear Regulatory Commission's internal mail system, this 21st day of July,1980:
Sheldon J. _Wolfe, Esq., Chairman
- Richard Lowerre, Esq.
Atomic Safety and Licensing Board Panel Asst. Attorney General for the U.S. Nuclear Regulatory Commission State of Texas Washington, DC 20555 P.O. Box 12548 Capitol Station Dr. E. Leonard Cheatum Austin, Texas 78711 Route 3, Box 350A Watkinsville, Georgia 30677 Hon. Jerry Sliva, Mayor City of Wallis, Texas 77485 Mr. Gustave A. Linenberger
- Atomic Safety and Licensing Board Panel Hon. Johri R. Mikeska U.S. Nuclear Regulatory Commission Austin County Judge Washington, DC 20S55 P.O. Box 310 Bellville, Texas 77418 Mr. John F. Doherty 4327 Alconbury Street Houston, Texas 77021 J. Gregory Copeland, Esq.
Mr. and Mrs. Robert S. Framson Baker & Botts 4822 Waynesboro Drive One Shell Plaza Houston, Texas 77035 Houston, Texas 77002 Mr. F. H. Potthoff, III Jack.Newman, Esq.
\\
7200 Shady Villa #110
-Lowenstein, Reis, Newman & Axelrad Houston, Texas 77055 1025 Connecticut Avenue, N.W.
Washington, DC 20037 D. Marrack 420 Mulberry Lane Carro Hinderstein Bellaire, Texas 77401 8739 Link Terrace Houston, Texas 77025
+
4 2-Texas Public Interest Margaret Bishop Research Group, Inc.
11418 Oak Spring c/o James Scott, Jr., Esq.
Houston, Texas 77043 13935 Ivymount Sugarland, Texas 77478 Brenda A. McCorkle 6140 Darnell Houston, Texas 770/4 J. Morgan Bishop 11418 Oak Spring Mr. Wayne Rentfro Houston, Texas 77043 P.O. Box 1335 Rosenberg, Texas 77471 Stephen A. Doggett, Esq.
Pollan, Nicholson & Doggett Rosemary N. Lemmer P.O. Box 592 11423 Oak Spring Rosenberg, Texas 77471 Houston, Texas 77043 Bryan L. Baker
,1923 Hawthorne Houston, Texas 77098 Robin Griffith Leotis Johnston 1034 Sally Ann 1407 Scenic Ridge Rosenberg, Texas 77471 Houston, Texas 77043 Elinore P. Cummings Atomic Safety and Licensing
- 926 Horace Mann Appeal Board Rosenberg, Texas 77471 U.S. Nuclear Regulatory Comission Washington, DC 20555 Atomic Safety and Licensing
- Board Panel U.S. Nuclear Regulatory Commission Mr. William Perrenad Washington, DC 20555 4070 Merrick Houston, TX 77025 Docketing and Service Section
- Office of the Secretary Carolina Conn U.S. Nuclear Regulatory Comission 1.414 Scenic Ridge Washington, DC 20555 Houston, Texas 77043 Mr. William J. Schuessler U.S. Nuclear Regulatory Commission 5810 Darnell Region IV Houston, Texas 77074' Office of Inspection and Enforcement 611 Ryan Plaza Drive The Honorable Ron Waters Suite 1000 State Representative, District 79 Arlington, Texas 76011 3620 Washington Avenue, No. 362 Houston, TX 77007
}/] (
QQ if
~
Stephenfl.Sohinki Counsel for NRC Staff
.