ML19323D849
| ML19323D849 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 05/20/1980 |
| From: | Sohinki S NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | DOHERTY, J.F. |
| References | |
| NUDOCS 8005220433 | |
| Download: ML19323D849 (25) | |
Text
. _ _ - _. _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
May 20, 1980 UNITED STATES OF AMERICA NUCLEAR REGULATORY t,3MMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 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 TWELFTH SET OF INTERR0GATORIES The NRC Staff responds as follows to John F. Doherty's twelfth set of inter-rogatories dated February 19, 1980. This pleading includes the Staff's response to approximatel tnirty of the interrogatories propounded to the Staff by Mr. Doherty on February 19.
By previous agreement with Mr. Doherty, the Staff will endeavor to file the balance of its responses on or before June 2,1980.
12-14 In NUREG-0401, p.14, there is a description of an event where fuel rods failed in 113 bundles and their number was "in excess of 200".
"The MSLRM during this transient did not alarm..." according to NUREG-0401, and "...this incident illustrates the insensitivity of the MSLRM..." (ibid.). How does the Office of Nuclear Reactor Regulation conclude the activity monitors "...alona with outer primary systems sensors" are "... adequate to give an early warning and allow a timely response to degrading fuel conditions."
(Staff Response to this Intervenor's Interog.1-14, P.17 and 18)?
(Note the event was in a BWR, smaller than ACNGS)
Response
The key to our response to Interrogatory 12-14-01 lies in the interpretation of degrading fuel conditions.
In the event referenced at Dresden, approximately 3.6% of the fuel rods in the 113 fuel assemblies are assumed to have failed. On O
. pe
' a core load basis of 724 fuel assemblies (732 assemblies in Allens Creek),
this gives a defect level of approximately 0.55%.
Furthemore, one has to consider the nature of the defect and the time frame involved in defect fomation.
In this case, the defects were produced by pellet / cladding inter-action (PCI).
Such defects are generally in the form of small tight cracks in the cladding which restrict the fission produce releases. Also, PCI defects do not occur instantaneously or even simultaneously and in this event, were observed to happen over a five hour span.
The overall event at Dresden is classified as one with unfavorable fuel performance but not one where gross fuel degradation has affected the safe operation of the piant.
The primary radiation sensors in a BWR are the Main Steam Line Radiation Monitor (MSLRM) and the Off Gas System Radiation Monitor (0GSRM), the latter at the steam jet air ejector. The features and limitations of these sensors are fully described in NUREG-0401.
For example, it is stated that the MSLRM would be insensitive to less than roughly 50 instantaneous failed rods and in such an event, would alarm within 7 seconds.
On the other hand, the OGSRM would be sensitive to a very few fuel rod failures but the signal would lag the event by 2-3 minutes.
Under the conditions described above for the Dresden event, the MSLRM would not have been expected to alarm and it didn't.
Near the end of the event when multiple failures were thought to have occurred in the last 30 minutes, the OGSRM alarmed on both high and high-high level. Thus, the simultaneous failure of a few fuel rods did alarm on the OGSRM and allow the operator to take pre-
- cautionary measures before the safety of the plant was impaired.
Thus, the i
1
combination of the OGSRM and the MSLRM appears to adequately monitor the coolant activity and gives an early warning before gross fuel degradation has occurred.
12-14 How, from the above (12-14-01, above) does the current MSLRM system contribute anything to the other systems toward detecting fuel failure?
Response
The main feature of the MSLRM system is to detect a large number'of simultaneous failures (~50 fuel rods) and to quickly alarm @-7 seconds) after such an event.
The operator can shut down the power operation before wide scale damage occurs.
12-14 Statement (4) of Group II on page D-3 of SER Supp#2 for ACNGS does not indicate that the systems can detect fuel failure suf-ficiently to permit prevention, but merely that the failure will be detected.
Is there any intent in the ACRS statement to mean ir, time to prevent fuel damage? If so, please inform by quoting any ACRS meeting on the topic or summarize and give a reference I might obtain through F0IA?
Resoonse Radiation monitoring systems discussed above depend upon the release of some fission product to activate their detection capability.
Thus, at this point, some fuel cladding fcilure has already occurred. There are no known instruments that could be practicably applied in LWR cores to detect incipient failures during operation, thus allowing shut down before failure.
However, incipient failure limits are under study in various research and development programs.
Operating experience, where failures have occurred, is also studied and procedures 6.
9
revised to eliminate these events in the future. An example of this is the Preconditioning Interim Operating Management Recommendation (PCIOMR) procedures developed by General Electric to minimize the incidence of failure due to pellet /claddir.g interaction (PCI).
Other fuel manufacturers have similar programs.
12-14 Does staff take the position a PCMA (or flow blockage accident) is so unlikely that it need not be considered in relation to the problem of rapid fuel failure detection?
(See Staff's reply to this Intervenor's Interog. #14 of Set #1 for more).
a/
If so, will staff then oppose any discussion of rapid fuel failure detection in the event of a flow blockage accident at the Construction License Hearing?
Response
A previous response in regard to fuel failure detection for BWR flow blockage has been filed in reply to Interrogatory 10-13.
As given above in our response to Interrogatory 12-14-01 and 12-14-02, the capabilities of the radiation monitoring systems are reviewed.
The MSLRM is expected to alarm 7 seconds after the simultaneous fctlure of at least 50 fuel rods.
The OGSRM is expected to alarm 2-3 minutes after the failure of a few fuel rods and should be able to detect a single rod failure. Therefore, the monitoring system as installed is expected to give an early warning of fuel degradation.
This will allow the operator to take precautionary measures before the safety of the plant is impaired.
P be t
ee 1
On the basis of the above response, we conclude that rapid fuel failure detection is available. We would not oppose any discussion of this matter at the hearing.
12-20 State some of the sources of " Conservatism" in the enthalpy limit guide as mentioned in your reply to this Intervenor's Interrog. 9-16, please.
Response
The conservatisms as mentioned in our response to Interrogatory 9-16 have been discussed in some detail in our responses to Interrogatories 8-10 and 9-26, especially in the latter. The conservatisms are based on two factors:
(1) the calculated enthalpy values are consistently well below the 170 cal /gm and 280 cal /gm limit levels and (2) the General Electric model used for these calculations doesn't take credit for all of the realistic core parameters.
A discussion of some of these input parameters was furnished in our response to Interrogatory 9-26.
In particular, the effect of including moderator feedback reduced the calculated peak enthalpy for the rod drop accident from 135 cal /gm to 35 cal /gm.
Using only the correct scram function, the calculated peak enthalpy was reduced from 135 cal /gm to 100 cal /gm. All of these enthalpy values are well below the fuel damage limit of 170 cal /gm.
12-20 Hai G.E., like Westinghouse, now developed a code to predict FGR in high burn-up fuel?
Response
Yes, General Electric has developed a fuel performance code to predict fission gas release in high burnup fuel. The code, called GESTR (Ref.1), is currently 1
under review by the staff.
The previously approved GE fuel performance code,
called GEGAP-III (Ref. 2), can also be used to predict fission gas release in high burnup fuel when modified by an NRC correction function (Ref. 3).
References :
1.
E. B. Johansson, G. A. Potts and R. A. Rand, "GESTR: A Model for the Prediction of GE BWR Fuel Rod Thermal / Mechanical Performance," General Electric Company Report NED0-23785, March 1978.
2.
"GEGAP-III: A Model for the Prediction of Pellet / Cladding Thermal Conductance in BWR Fuel Rods," General Electric Company Report NEDO-20181, November 1973.
3.
R. O. Meyer, C. E. Beyer and J. C. Voglewede, " Fission Gas Release from Fuels at High Burnup," U.S. Nuclear Regulatory Commission Report NUREG-0418, March 1978.
12-20 If the answer to 12-20-02 is "yes" does the G. E. model assume release rates are the same at high burn-up as at low?
Response
The previously approved GE fuel performance code, GEGAP-III, utilizes a fission gas release model which is a function of temperature only. At a given temperature, the release rate is the same at high burnup as at low burnup.
This model must be modified with the previously mentioned NRC correction in order to consider burnup effects. The new code, GESTR, utilizes a fission gas release model that is a function of both temperature and burnup independent of any correction.
Therefore, both the new GESTR code and the older GEGAP-III code with the NRC correction assume release rates that are different at high and, low burnup.
12-20-06
.By)how much does fuel temperature decrease with burn-up between (a 0 -10,000 mwd / tonne U.; (b) 10-20,000 mwd / tonne U. (c) 20-30,000 mwd / tonne U.; (d) 30-45,000 mwd / tonne U.?
i I
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. Response Because fuel temperatures depend on a large number of specific conditions (e.g., fuel design, operating conditions, reshuffling scheme), it is difficult to quantify fuel temperature decrease with burnup. However, a general indication of how fuel temperatures vary with burnup can be found in Figure 1.
This figure shows GAPCON-2 predictions of fuel avarage temperature as a function of burnup for several different constant power levels.
Power maneuvering, fuel depletion and other effects were not considered in these calculations.
It can be seen from the figure that the simplified constant power burnup history does not lead to monotonically decreasing fuel temperatures.
12-20 Is GAPCON a G.E. code? If "yes" when approved by NRC?
Response
No, GAPCON is not a General Electric code.
The name GAPCON refers to a series of fuel performance codes originally developed (Ref.1) by Hanford Engineering Development Laboratory and subsequently modified (Refs. 2-5) by Battelle Pacific Northwest Laboratories for the Directorate of Licensing of the U.S. Atomic Energy Commission and later for the Nuclear Regulatory Commission's Office of Nuclear Reactor Regulation. The GAPCON codes are used by the staff to audit vendor safety analyses.
Although fuel performance codes similar to the GAPCON series have been approved for use in safety analyses, the GAPCON series itself has never been used by a licensee.
References:
1.
G. R. Horn, F. E. Panisko, Users's Guide for GAPCON: A Computer Program to Predict Fuel-to-Cladding Heat Transfer Coefficients in Oxide Fuel Pins, Hanford Engineering Development Laboratory Report TME72-128, September 1972.
i.
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2.
C. E. Beyer, C. R. Hann, D. D. Lanning, F. E. Panisko, and L. J. Parchen, "GAPCON-THERMAL-2: A Computer Program for Calculating the Thermal Behavior of an Oxide Fuel Rod," Battelle Pacific Northwest Laboratory Report BNHL-1898, November 1975.
3.
C. E. Beyer, C. R. Hann, D. D. Lanning, F. E. Panisko, and L. J. Parchen, User's Guide for GAPCON-THERMAL-2: A Computer Program for Calculating the Thermal Behavior of an Oxide Fuel Rod," Battelle Pacific Northwest Laboratories Report BNWL-1897, November 1975.
4.
D. D. Lanning, C. L. Mohr, F. E. Panisko and K. B. Stewart, "GAPCON-THERMAL-3 Code Description," Battelle Pacific Northweat Laboratories Report PNL-2434, January 1978.
5.
D. D. Lanning, F. E. Panisko and C. L. Mohr, "GAPCON-THERMAL-3 Verification end Comparison to In-Reactor Data, "Battelle Pacific Northwest Laboratories Report PNL-2435, September 1978.
12-20 How accurate are measurements of fuel burn-up at more than 30,000 mwd / tonne U.? Is the accuracy at 30,000 mwd / tonne U. considered as good as at 40,000 mwd / tonne U.?
Response.
The accuracy of fuel burnup measurements depends upon the method used. The precision of analyses for var'.us techniques are listed in Table 1 of the ASTM
- 219-69 standard. The precision for total heavy element atom percent fission, F, ranges from 0.9% for the ND-148 mass spectrometry method to 6.0% for the T
CS-137 colorimetry method. These methods require a destructive examination of the irradiated fuel and are, in general, only used in the development and verification of computer codes fur burnup predictions.
With these computer codes, the fuel burnup for Allens Creek throughout each cycle and for end-of-life conditions is calculated. The uncertainty of power distribution and consequently burnup within a fuel assembly is + 5%. The 0
.ao C___________
. predicted burnup values for Allens Creet fuel should be accurate to within
+ 10%. This accuracy should be expected for any burnup level.
12-20 Does Staff have data supporting a statement on Page 26 of NUREG-0418 that, "...some fuels in BWRs experience a marked decrease in gap conductance with burn-up as a result of helium dilution in the gap",
a/ Please site supporting data for this, b/ Is the statement true for all fuel enrichments for ACNGS?
c/ For how great exposure does this statement hold?
d/ For what maximum design linear heat generation rate range?
Response
The statement on page 26 of NUREG-0418 refers to the susceptibility of the BWR fuel design to fill gas dilution with fission gases.
All BWR fuel rods are backfilled with helium during the manufacturing process.
During operation, the released fission gases (notably Xenon and Krypton) mix with the original helium fill gas in the fuel-to-cladding gap. The thermal conductivity of the resulting mixture is lower than that of pure helium.
GAPCON-2 predictions of this effect are shown in Figure 2.
The dilution process can be verified by post-irradiation chemical analysis of the gas mixture.
Typical analytical results of this type are found in Refs.1 & 2.
The helium dilution process occurs to some degree in all LWR fuels, for all enrichments, exposures and heat generation rates.
Because fuel-to-cladding gap conductance depends not only on the thermal conductivity of the gas mixture
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but also on the gap dimensions, the gap conductance may increase at some burnup levels. However, the BWR fuel design, with its low fill gas pressure (relative to PWRs) and its low plenum volume (relative to EBR-II fuels) is particularly vulnerable to the dilution process.
The newest BWR fuel design (placed in service after the publication of NUREG-0418) uses a higher (3 atmosphere as opposed to 1 atmosphere) fill gas pressure to minimuze this problem. The GAPCON-2 predicted gap conductance for this new design is shown in Figure 2.
The same design was also used in Figures 3 and 4 This newer higher pressure fuel that is less susceptible to this dilution phenomenon will be used in ACNGS.
References:
1.
P. Knudsen and C. Bagger, " Power Ramp and Fission Gas Performance of Fuel Pins M20-1B, M2-2B and T9-3B," Risd National Laboratory Report Risd-M-2151, December 1978.
2.
C. Bagger, H. Carlsen and P. r udsen, " Details of Design, Irradiation and Fission Gas Release for the Danish UO -Zr Irradiation Test 0ZZ," Ris6 2
National Laboratory Report Rist-M-2152, December 1978.
12-20 What is the fuel operating temperature of an LMFBR?
Response
The operating temperature for fuel in a LMFBR is represented in the enclosed figure (Figure 5) from the Clinch River PSAR.
Fuel centerline temperatures range from 2800 F to 4200 F while the fuel surface temperatures range from 1375 F to 1950*F.
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. 12-20 In a letter to D. F. Ross (NRC) from G. G. Sherwood (GE), it states, "For 8x8 bundles, the NRC exposure dependent FGRM does not result in any calculated perforation (of fuel rods) even at bundle average planar exposures of 30,000 ffWd/St." Assuming Mr. Sherwood correct, are there any calculations of perforation when the exposures are raised to 45,000 mwd /t an exposure proposed for ACNGS, on Page 4.2-10 of the PSAR?
Response
The burnups cited in Interrogatory 12-20-12 are for two different designated parameters.
The 30,000 mwd /St is listed correctly as the bundle (fuel assembly) average planar exposure. The value cited on page 4.2-10 in the Allens Creek PSAR of 45,000 mwd /St is labeled as maximum local exposure.
In the local case, the value is the maximum nodal exposure within a fuel assembly which corresponds to a 6-inch axial segment.
In general, the maximum local exposure exceeds the bundle average exposure by 10,000 to 15,000 mwd /St at end-of-life.
Thus, while the two designated burnup levels are numerically different, the given values correspond to an approximate identical operating situation.
To study fuel performance at higher burnups (not proposed in the Allens Creek PSAR), General Electric is participating in a combination of DOE-EPRI programs with lead test assemblies at Monticello and Peach Bottom 2.
These programs will furnish test results in 1983 to 1985.
12-20 Is NRC currently developing a generic PCI mode??
a/ If so, when is its completion expected?
12 -
Response
As part of a cooperative effort between the Chalk River Nuclear Laboratory, operated by Atomic Energy of Canada, Ltd., and Battelle Pacific Northwest Laboratories (a consultant to NRC), a PCI model, called PROFIT, was developed.
The cooperative study and PROFIT PCI model are described in a report entitled, "PCI Fuel Failure Analysis: A Report on a Cooperative Program Undertaken by Pacific Horthwest Laboratory and Chalk River Nuclear Laboratories," NUREG/CR-1163, December 1979.
The report is available through the GP0 Sales Program, Division of Technical Information and Document Control, NRC.
Development work on the PROFIT model is continuing at Battelle during the carrent fiscal year.
When the model is finalized, NRC will use it in performing audit calculations for transient and accident analysis.
12-20-14_- Have all G. E. BWRs when fuel was burned-up beyond 30,000 mwd /
tonne U. been required to lower the Maximum Average Planar Linear Heat Generation Rate, as was Monticello in Amendment #42 to its OL recently?
a/ If not, was Monticello an exception to the rule?
Response
It is assumed that all BWRs will be required to lower their MAPLHGR values for burnups beyond 30,000 mwd /t.
Enclosed are Figure 3.2.1-8 (Figure 6) from the Brunswick Unit 2 Technical Specifications and Figure 3.5.1.F (Figure 7) from Peach Bottom 2.
The general trend, as shown in the figures, applies to all operating BWRs. The allowable value for MAPLHGR peaks between 10,000 and
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, 15.000 mwd /t and then diminishes at higher burnups.
Since 30,000 mwd /t was assumed to be the upper level for normal burnups, the values for MAPLHGR were only calculated and plotted to this limit.
However, the trend is clearly established and lower MAPLHGR values would be expected at burnups greater than 30,000 mwd /t as shown for Peach Bottom 2.
12-20-1E - What is the general effect of crud deposition on the cladding surface of a BWR fuel clad on the inside?
Response
Crud deposition, as seen on normal operating fuel rods, is found on the outside surface of the cladding.
In general, crud is an oxide mixture of various metallic cations found in the coolant, namely, iron, nickel and chromium.
Such a layer has a lower thermal conductivity than the metallic cladding, thus restricts thermal conductance to the coolant. When crud layers become significantly thick ( 2-3 mils), spalling occurs, thus removing the crud and returning the thermal conduction path to its normal condition.
The interrogatory questions the formation of crud on the inside of the cladding. None would be expected on the inside since the material for crud formation comes from the coolant.
If the term crud is being applied to the U-Zr-fission product interface layer found on the inside surface, the general effect is not expected to be as significant. Gap conductance is enhanced by the solid-solid bonding as opposed to conductance across a gas filled gap.
Some of the fission products, cesium and iodine, have been identified as agents for stress corrosion cracking,, but
~ 'his process is more complicated than their mere presence.
Such behavior is t
i l
part of the PCI considerations.
. 12-20 In the AEC's evaluation of GEGAP-III, titled "Supp#1 to Technical Report on Densification of G.E. Reactor Fuels" 12/72, the report mentions wel; t5aracterized experiments that predict equiaxed grain growth radii conservatively.
Does the NRC take this position with regard to equiaxad grain growth prediction today? If not, what new report have shaken confidence in the predictions?
Response
The staff's evaluation of " Supplement 1 to the Technical Report on Densification of General Electric Reactor Fuels," (Ref.1) dated December la,1973, discusses the use of grain growth kinetics in the GE densification model. Although the model predictions were in good agreement with Halden reactor data (Refs. 2 and 3), we concluded that (1) the GE model did not correctly nor conservatively predict the densification behavior of high-density, stable fuels similar to the GE product line and (2) more appropriate coefficients in the densification model, based on grain growth kinetics, would be difficult to obtain. As a result,two staff-imposed restrictions were placed on the GE densification model.
The first requires the predicted maximum density to occur no later than 4,000 mwd /t. The second requires the predicted maximum density to be related to out-of-reactor resintering data from production fuel. We continue to believe that in-reactor grain growth kinetics are very difficult to determine and that the staff-imposed restrictions are appropriate for high-density, stable fuels similar to the GE product line.
Additional and more recent information on this position can be found in Refs. 4 and 5.
References:
1.
"GEGAP-III: A Model for the Frediction of Pellet-Cladding Thermal Conductance in BWR Fuel Rods," General Electric Company Report NED0-20181, November 1973.
.m
. 2.
U.S. Atomic Energy Comnission Regulatory Staff, " Supplement 1 to the Technical Report on Densification of General Electric Reactor Fuels,"
December 14,1973 (Reproduced in NUREG-0085.
See Reference 4 below).
3.
A. Hanevik, P. Arnesen and K. D. Knudsen, "In-Reactor Measurements of Fuel Stack Shortening," Paper Nr 89 presented at BNES Nuclear Fuel Performance Conference, London, October 15-19, 1973.
4.
R. O. Meyer, "The Analysis of Fuel Densification," U.S. Nuclear Regulatory Commission Report NUREG-0085, July 1973.
5.
Regulatory Guide 1.126, "An Acceptable Model and Related Statistical Methods for the Analysis of Fuel Densification," U.S. Nuclear Regulatory Commission, March 1977.
Doherty Contention #31 Flow induced Vibration and LPRM degradation 12-31 What is the NRC document with the diagram of the fix for flow induced vibration of the fuel assemblies? Can you direct this Intervenor to it, or mail a copy of the document or the part showing the fix?
Response
When the general problem arose in 1975 with the vibrating instrument tubes I'PRM) impacting fuel channels, almost all of the BWRs had bypass flow holes in the core support plate. The fix at that time was to mechanically plug the core support plate holes. A number of similar General Electric topical reports were issued on the subject. A typical one is:
" James A. Fitzpatrick Nuclear Power Plant Channel Inspection and Safety Analysis with Bypass Holes Plugged," NED0-21166, January 1976.
For,BWR plants under construction or in the planning stage, the bypass holes in the core support plate were eliminated from the design. Thus, plugging oe n
- devices were not required and this is the case with Allens Creek. Attached are two figures which show the bypass leakage paths before and after the fix.
In Figure 8, the pre-1975 design is shown. The bypass flow path that initiated LPRM vibration is designated as No. 9 - Core Support Plate Moles.
Note also in this figure that the Lower Tie Plate has no holes.
The fix referred to above was a simple plugging of the core support plate with a metal plug and drilling a few smai holes into the lower tie plate.
Figure 9 shows the current BWR design watch will be used in Allens Creek.
The core support plate has no holes for bypass flow and the lower tie plate on the fuel assembly is drilled to allow some bypass flow.
These are the only changes between the two designs and operating experience to date indicates that the changes have been effective in eliminating instrument tube vibration.
12-31 On page D-IV of NUREG-0572, (ACRS Report on LERs) it states, "Other failures resulting from vibration include... radiation monitor failures." Does ACRS or any other NRC body have any planned research or meetings planned where the topic of these failures will be taken up?
Response
The ACRS Subcommittee on LJRs met on April 22, 1980, but the subject of flow induced vibration was not ditcussed.
Since the ACRS subcommittee, in September 1979,~ recommended further studies in this area, one can assume that the subject matter will be discussed at some future meeting.
In addition to the ACRS, the newly formed NRC Office for Analysis and Evaluation of Operational Data will be involved with such matters.
l l
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NOTE rt nerH[R AL F(f( L SurPOHis ARE wt LO[D IN101Ht CORL SurrORT PLAlt iOH THESt DUP 4DLES.
FAT H NOUD[HS 1,7.5 AND 7 DO p
NOT E XIST.
. CHANNEL i
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LOWER TIE PL ATE _
6 CORE surrORT PLATE FUE L SUPPORT l
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IN CORE GulOE TUBE g
I 1 ~ SHROUD CONTROL ROD
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GUIDE TUBE
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- 1. CONTROL ROD GUIDE TUBE. FUEL SUPPORT CASTING
- 2. CONTROL ROD GUIDE TUBE CORE SUPPORT PLATE 3 CORE SUPPORT PL ATE INCOR5 GUlOE TUBE
- 4. CORE SUPPORT PL ATE. SHROUD l
5 CONTROL ROD GUIDE TUBE DRIVE HOUSING G. FUEL SUPPORT.LOWEH TIE PLATE
- 7. CONTROL ROD DRIVE COOLING WATER y
- 8. CHANNEL. LOWER Tif PL ATE 9.
CORE SUPPORT PLATE HOLES d
5
. CONTROL ROD DRIVE HOUSING 1
__ 4 Tisure 5-3.
Schematic of Reactor Assetbly Showing the Leakage Flow Paths Fly e i O.2-31-CO
tJOTE: PE RIPHE R AL F UE'l SUri OR TS ARE WE LDID INTD 1HE CORE L-
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SUPPORT PLATE FOR THLSE BUNO LE S, PAT H NUMBE R$ 1, g CHANNEL
- 2. 5 ANO 7 DO NOT EXIST.
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TIE PL ATE -
Coele SUIS ORT PLATE -
FUEL SUI' PORT
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IN CORE GUIDE TUBE I
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SHROUD CONTROL ROD g
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- 1. CONTROL ROD GulOE TUBE FUE L SUPPORT
- 2. CONTROL ROD GulOE TUBE CORE SurPORT PLATE
- 3. CORE SUPPORT PLATE INCORE GusOE TUBE I
- 4. CORE SUPPORT PL ATE SHROUD
- 5. CONTROL ROD GUiOE TUBE DRIVE HOUSING
- 6. FUE L SUPPORT LOWER TIE PLATE 7
- 7. CONTROL ROD DRIVE COOLING WATER
- 8. CHANNE L LOWE R TIE PLATE
- 9. ALTERNATE FLOW PATH HOLES e
pe CONTROLslOI)
DRIVE HOUSING I
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Figure 4-1.
Schematic of iteactor Assembly Sliowing the liypass I' low Paths p, y,, g g C
(19 08)
. 12-25 Is xenon tagging being considered by NRC for fuel loads as suggested on Pg. 20 of NUREG-0401 for ACNGS or other nuclear power plants?
Response
Xenon tagging of fuel rods is expected to find regular application only in the sodium cooled reactors (LMFBR) where fuel assemblies cannot easily be visually inspected or checked for fuel failures.
The added cost of tagging is justified if the operator must locate a specific leaking assembly under liquid sodium.
In LWRs, detection and identification of suspect or leaking assemblies can easily be performed, sometimes in-core as with sipping heads in BWRs or with sipping equipment in the spent fuel pool.
Therefore, while xenon tagging might narrow down a sector for inspection, any benefits of the design would be outweighed by the added cost.
12-Mc In the AEC Evaluation of GEGAP-III (Supp-I to Technical Report on Densification of GE Reactor Fuels 12/72) it states the den-sification model indicate maximum density occurs by 4,000 mwd /
tonne U.
Is this still believe correct? What modifications in this behalf or finding have taken place since publication?
Response
Yes, it is our opinion that maximum in-reactor densification in oxide fuel occurs by 4,000 mwd /t.
The burnup at which the densification process is essentially complete is a function of several parameters, including the fabrication technique.
Densification occurs more rapidly and to a greater 1
degree in the older low-density and unstable fuels. Additional information, published since the December 14, 1973 evaluation, (Ref. 1) can be found in Refs. 2-4.
These references confirm the early-in-life densification peak.
References:
1.
U.S. Atomic Energy Commission Regulatory Staff, " Supplement 1 to the Technical Report on Densification of General Electric Reactor Fuels,"
December 14, 1973 (Reporduced in NUREG-0085. See Reference below).
2.
A. Hanevik, P. Arnesen and K. D. Knudsen, "In-Reactor Measurements of Fuel Stack Shortening," Paper No. 89 oresented at BNES Nuclear Fuel Performance Conference, London October 15-19, 1973.
3.
R. O. Meyer, "The Analysis of Fuel Densification," U.S. Nuclear Regulatory Commission Report NUREG-0085, July 1976.
4.
Regulatory Guide 1.126, "An Acceptable Model and Related Statistical Methods for the Analysis of Fuel Densification," U.S. Nuclear Regulatory Commission, March 1977.
12-32 Was a meeting between an NRC group and G. E. titled "Two loop test apparatus Implications for ECCS Models, ECCS Model Errors" scheduled for 5/24/79 in Phillips Bldg., ever held? How can I get more information on it? Can you give me a summary of its outcome?
Response
Yes, the meeting was held on 5-24-79 in the Phillips Building, Bethesda, Maryland. A summary of the meeting dated June 12, 1979, is enclosed.
12-32 How do parallel channel effects (PCE) effect the vaporization rate of the ECCS in a BWR?
Response
No effect is calculated. With a supply of water in each channel, the vaporization rate calculated within each channel is a function of the heat i
generation rate and the water temperature within the channel.
6.
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_ - _ _ 12-32 Has the problem described in this contention been part of any other licensing proceeding for a BWR?
Response
Yes, it has been a part of the proceeding for Black Fox Station, Units 1 and 2.
12-32 Has it been part of any ACRS reports or meeting other than the 3/8 to 3/10/79 ACRS meeting (#227)?
Response
No.
12-32 Has a Supplement to NUREG 0528, SER to Wm. Zimmer Plant, Unit 1, been published? (This unit first raised the Contention #32 issue).
Response
No.
Contention #40 Assumed fission product release is incredible 12-40 (You may wish to refer to my Interog 8-6).
Language on Page 11 of NUREG-0588, " Population Dose and Health Impact of TMI-2", seems to indicate there is considerable doubt of the exact amount of each source term released at that accident and on which must of Contention 40 is based.
a/ Is there any known release amount measurement of the quantity of each material (that is element or isotope) released from TMI-2, in the March 28, 1979, event.
b/ Please indicate where the information is available.
j (Note:
I am not interested in total alpha, beta or gamma, I want to know the amount of curies of each substance.)
. ee
Response
The radioactivity that was released to the aquatic environment is described in the licensee's Radioactive Effluent Monitoring Release Report for the period January 1,1979, through June 30, 1979, as well as in draft NUREG-0598,
" Release of Radionuclides into the Susquehanna River from Three Mile Island Nuclear Station during the Period of 3/28/79 - 5/11/79." The releases to the atmosphere are described in an NRC memorandum from L. Barrett to Distribution dated April 12,1979, " Preliminary Estimates of Radioactivity Releases from Three Mile Island" and in an NRC memorandum from R. C. DeYoung to W. Kreger dated October 3,1979 " Calculated Offsite Iodine-131 Air Concentrations from Three Mile Island." All of the above documents are available in the NRC Public Document Room.
12-40 The Kemeny Commission report seems to point to a lack of testing requirements (page 30) to verify auxiliary building filter effectiveness.
a/
Is NRC considering establishing these testing requirements?
b/ Wouid changes in requirements apply to the ACNGS design?
Response
a.
U.S.NRC Regulatory Guides 1.52, " Design, Testing, and Maintenance Criteria For Post Accident Engineered-Safety-Feature Atmospheric Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants,"
and 1.140, " Design, Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," for Engineered-Safety-Feature (ESF) and non-ESF
. ee
. 4 ventilation systems respectively, both recommend inplace testing.
Implementation of this guidance in the form of requirements in the station technical specifications is currently applied only to ESF systems.
Consideration is presently being given to implementing the guidance for non-ESF systems as well.
b.
If established, such changes will apply to plants under operating license review and may also be implemented for operating reactors.
i 12-40 What has been concluded as the reason for pool filter performance i
in filtering I-131 (mainly in methyl iodide) during the TMI-2 accident?
Response _
Data obtained from an analysis of TMI-2 charcoals and an independent study by the Rogovin Commission concluded that the auxiliary and fuel-handling building exhaust filter systems installed at the time of the accident provided an overall decontamination factor of approximately 9.5 (equivalent to an 89.5%
efficiency) for radiofodine.
Therefore, the charcoal did in fact remove much of the radiciodine from the exhausts. The somewhat degraded performance of the filter trains is attributed to:
(a) degradation of the charcoal due to continuous operation of the systems (fuel-handling building) for slightly greater than a i
year, and potential exposure to significant amounts of fumes and aerosols generated during final painting and cleanup of the fuel-handling and auxiliary buildings, (b) use of carbon which would not meet current regulatory requirements (in 'the fuel-handling building exhaust system), and (c) lack of inpalce testing requirements for the fuel-handling building systems.
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. 12-40 Does staff take the position that if the charcoal filters mentioned about(in 12-40-03) met the regulatory requirements of today, the I-131 would not have passed through the filter?
Response
Since charcoal filtration is not absolute, it is expected that some iodine would have been released under the best of circumstances.
With periodic inplace testing and the use of carbon which meets current regulatory require-ments in the fuel-handling building ventilation exhaust system, it is felt that iodine releases could have been lower by a factor of approximately 5.
12-T TEXPIRG Contention 30 Has G.E. responded to the request of 0. D. Parr (NRC) to Sherwood (GE) of 4/19/79 (NRC #7905080114) with regard to supplying additional occupational exposure for the v3dwaste system as described in NED0-21,059 "BWR Radioactive Wast? Treatment Systems"?
If so, pelase supply a summary or some information c1 that.
Incidently applicant reported not being in possession of NED0-21,059.
Response
ilo. We will review the response according to Regulatory Guide 8.8 criteria when the response is received.
Dated at Bethesda, Maryland, this 20th day of May, 1980.
6.
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e e,
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of HOUSTON LIGHTING & POWER COMPANY 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.
D$~ $4N'f}Zmy Calvin W. Moon Subscribed and sworn to before me this, 6 day of h.%.
1980.
bibl. A b, Notary Public
/-
~
My Comission expires:
4 / /MV y
u 6.
4
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 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. D0HERTY'S TWELFTH SET OF INTERROGATORIES" and " AFFIDAVIT OF CALVIN W. MOON" 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 by deposit in the Nuclear Regulatory Commission internal mail system, this 20th day of May,1980:
Sheldon J. Wolfe, Esq., Chairman
- Richard Lowerre, Esq.
Atemic 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. John R. Mikeska U.S. Nuclear Regulatory Commission Austin County Judge Washington, DC 20$55 P.O. Box 310 Bellville, Texas 77418 R. Gordon Gooch,.Esq.
Baker & Botts Mr. John F. Doherty 1701 Pennsylvania Avenue, N.W.
4327 Alconbury Street Washington, DC 20006 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 Jack Newman, Esq.
~
Mr. F. H. Potthoff, III 1814 Pine Village Lowenstein, Reis, Newman & Axelrad Houston, Texas 77080 1025 Connecticut Avenue, N.W.
Washi69 ton, DC 20037 D. Marrack 420 Mulberry Lane Carro Hinderstein Bellaire, Texas 77401 8739 Link Terrace Houston, Texas 77025
=0*
3 -- -
Texas Public Interest Margaret Bishop Research Group, Inc.
11418 Oak Spring c/o James Scott, Jr., Esq.
Houston, Texas 77043 8302 Albacore Houston, Texas 77074 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. Lammer P.O. Box 592 11423 Oak Spriag 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 Conmission Washington, DC 20555 Mrs. Connie Wilson 11427 Oak Spring Atomic Safety and Licensing
- Houston, Texas 77043 Board Panel U.S. Nuclear Regulatory Commission Mr. '4111iam Perrenod j
Washington, DC 20555 4070 Merrick Houston, TX 77025 Docketing and Service Section
- Office of the Secretary Carolina Conn U.S. Nuclear Regulatory Commission 1414 Scenic Ridge Washington, DC 20555 Houston, Texas 77043 Mr. William J. Schuessler 5810 Darnell Houston, Texas 77074 The Honorable Ron Waters AA State Representative, District 79
/M h
3620 Washington Avenue, No. 362 StMhefM.Sohinki Houston, TX 77007 Counsel for NRC Staff
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