ML20049H898
| ML20049H898 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 02/23/1982 |
| From: | Gilmore J CITIZENS FOR FAIR UTILITY REGULATION |
| To: | Atomic Safety and Licensing Board Panel |
| Shared Package | |
| ML20049H897 | List: |
| References | |
| NUDOCS 8203040450 | |
| Download: ML20049H898 (16) | |
Text
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A UNITED STATES OF A.'.! ERICA
<j NUCLEAR REGULATORY COMMISSION f'
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
TEXAS UTILITIES GENERATING
)
Docket Nos. 50-445 COMPANY, et al, 50-446 (Comanche Peak Steam Electric
)
Station, Units I and 2
)
(Application for
)
Operating Licenses)
MOTION FOR VOLUNTARY WITHDRAWAL OF CONTENTIONS TWO, THREE. FIVE AND SEVEN BY CFUR COMES NOW, Citizens for Fair Utility Regulation, hereinafter "CFUR", and file this its Motion for Voluntary Withdrawal of Contentions Two, Three, Five and Seven and for grounds would show this Board as follows:
I.
CFUR no longer has the financial capacity to conduct a meaningful it,ter-vention with regard to its remaining contentions. In that CFUR has no funds to retain further legal aid or further expert engineering or other professional aid feels that it is necessary to withdraw from the proceeding as an active party i:
intervenor.
II.
However CFUR makes such withdrawal without prejudice to the stipulations it has reached with Applicant regarding CFUR contentions One and Nine and further without prejudice to CASE's retention of and further litigation of Contention Five.
CFUR fully expects the stipulations reached between CFUR and Applicant with regard to said contentions One and Nine to be made a part of and be retained in the FSAR.
8203040450 820223 PDR A00CK 05000445 6
=
J This withdrawal is further made with the understanding that the NRC Staff has 9
j required as a condition of the Applicant's operating license that Applicant utilize grmmd water for CPSES only during peak water usage periods as stated in the Environmental Impact Statement for CPSES. (See FES, Summary and Conclusions,
- 9. c. vi). Such requirement resolved CFUR's concerns in its Contention Eight which was withdrawn as a result of such requirement.
III.
Furthermore CFUR alleges that Contentions Two, Three and Seven remain 4
valid, unresolved issues which must be resolved by the Applicant prior to Applicant's receiving an operating license. Accordingly CFUR urges this Bored to investigate and adopt CFUR's said three contentions in order to insure the safety of the public affected by Comanche Peak Nuclear Power Plant (also known as Comanche Peak Steam Electric Station). In furtherance of such urgings by CFUR, certain state-i ments of concerns and details of such facts as CFUR has been able to uncover with its meager resources are being attached to this pleading as Exhibits 1 and 2.
WHEREFORE, PREMISES CONSIDERED, CFUR respectfully prays that it be allowed to voluntarily withdraw its status as an Intervenor party and that 4
i this Board, rather than dismissing CFUR's Contentions Two, Three and Seven, adopt said contentions as their own.
1 Respectfully submitted, i
I t
J. MARSHALL GILMORE, Member Citizens For Fair Utility Regulation e
a b
g.9 h
e 1
ENHIBIT 1 CONT'ENTIONS TWO AND THREE STATEMENT No attempt is made to categorize the following problems according to the respective contentions.
The following problem areas are addressed:
1.
Lack of Proper Verification Use of Unacceptable Equation a.
b.
Reduction in Conservatism c.
Hafnium Centrol Rods 2.
Deletion of Boron Injection Tank 3.
LOFT and Semiscale Tests 4.
Omission of Safety Evaluation Reports for Topical Reports / Computer Codes.
5.
Human Factors I.
LACK OF PROPER VERIFICATION A.
Use of Unacceptable Equation In the NRC Staff Safety Evaluation of WCAP-S720 (Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations)I an empirically derived equation is identified which does not conform to known physical laws (the equation predicts a threshold where none could physically exist).
The Staff's accommodation of this situation was to reason as follows: Since the subject equation had been utilized in the derivation of fission gas release and IPages 6 & 7 of Enclosure to " Staff Evaluation of WCAP-S720",
Feb. ' 9, 197 9.
r
other models, the elimination of the equation or the application of limits on gap size dimensions would necessitate the rederivation of a number of models. Westinghouse presented an argument that the equation was utilized only in non-limiting areas of consideration in the subject report. So the Staff accepted the use of the equation with some restrictions and added that the use of the equation without restrictions in future license applications would not be acceptable.
The Staff's reasoning is faulty for two reasons. First, there was no l
attempt to determine whether or not the subject equation had any impact on the limiting area of concern. The fact that the equation models a non-limiting area of consideration does not in itself establish that the limiting area of concern would not be impacted to some extent. In this regard, not only was this detrmination absent for the subject report, it v as absent for the number of models which would require rederivation if the use of the equation was not accepted.
The second reason is that, as described in the Staff Safety Evaluation, the data used to estimate the constants of the equation were derived from experiments on eleven test items including both helium and fission gas-filled rods with a large range of cold gap sizes. The Staff notes that given the same data base, it would be difficult to independently reproduce the constants used in the equation.
None of the usual statistical measures concerning the use of the estimated constants was presented. But it is obvious from the description of the data base that no reasonable degree of confidence can be placed on those constants so estimated.
The use of such divergent test item configurations when only eleven items were tested coupled with a raw data fit that would be difficult to independently reproduce does not p'rovide a sufficiently adequate data base to draw hardly any kind of conclusion.
a Based on the,above consider'ations, no basis exists for a claim of conservatism as required by the regulations. Nor can any claim regarding a reliable "best estimate" be made with an adequate basis to support the position.
B.
Reduction in Conservation The "NRC Staff Safety Evaluation of the Request for Removal of Conservatism in the Westinghouse PAD Computer Code", is a lengthy tentative rational for allowing acceptance of a code which omits conservatisms not specifically required by law.
Discussions of uncertainties associated with (1) manufacturing variations, (2) comparisions of computer code predicted values to data described as: " Westinghouse believes the experimental fuel temperature data shown in Ref.11 to be representative of their standard product line" (p.19), and (3) comparisions,of Westinghouse code uncertainties with those of other computer codes abound. An analysis of fuel rod behavior under load following operating conditions including fuel rod bowing, pellet / cladding inter-action, waterlogging, fuel assembly fretting and wear at grid contact locations and baffle plate joints, and guide thimble wear to determine uncertainties between each of the 50,952 rods in a Comanche Peak reactor was ignored. Nevertheless, this is the typical manner in which the Applicant is proposing to utilize the plant. The history of a fuel rod during accident conditions will depend on the history of the particular fuel rod previous to the accident initiation. The listed conditions (as well as chemical uncertainties) could each individually substantially contribute to the history of a rod and could synergistically have a much larger impact. Localized effects, during 1
load following with the above uncertainties, could cause some fuel rods or groups of fuel rods to exhibit significant or different characteristics from that presented in
{
the report'. The computer code solution technique is not capable of taking into 2 Safety Evaluation of WCAP-8720", Feb. 9,1979 5-
consideratic.n localized conditions and it is necessary to allow for these specific uncertainties for such localized conditions.
Allowances for uncertainties due to steam generator tubes containing defects is also ignored. The events at the R. E. Ginna plant wherein the Applicant had concluded that the plant was able to be safely operated prior to subsequent findings of serious inadequacies demonstrates that present traditional tests such as eddy current tests are inadequate to determine the servicability of steam generators.
Yet no allowance for these uncertainties is addressed in the subject report.
Uncertainties also exist due to the inability of the Staff review process to discover errors in programs submitted to it. On March 23, 1978 Westinghouse informed the Staff that they had discovered as error in the LOCTA program and the SATAN program which had the effect of causing the metal water reaction heat 4
release to be cne-half of what it should be. The LOCTA program was formally accepted for use by the NRC Staff on May 30, 1975. The SATAN program was formally accepted for use by the NRC Staff on May 30, 1975. By letter dated December 22, 1981, the NRC was informed of an error in the ECCS analyses of four Westinghouse designed plants which are currently undergoing OL review and one currently operating plant which involved use of a single failure assumption not representing the most limiting condition. The report states:
l "The overall conservatism in the calculation of fuel temperatures and the overall conservatism in the LOCA l
analysis have not been rigorously demonstrated...
i The impact of the proposed modification on the overall analyses has not been addressed. As a result, we are 2
unable to consider the overall conservatism in the LOCA analysis as a bsis for the removal of the 65 F model uncertaint.y in PAD-3. 3".
Failure to take into consideration the above uncertainties cause the subject 6
report to be inadequate as a basis as welh a -
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]
C.
Hafnium Control Rods There has been no applicable experience gained with the use of hafnium control rods to be used in Comanche Peak. The use of hafnium control materials in the' Naval Reactor is not relevant because the hafnium used in such applications is in a different configuration and is unclad. The few instances when hafnium control material has been experimented with in PWR's is also not relevant since cruciform shaped rods were used that are non prototypic of the proposed usage.
Both Westinghouse and Brookhaven have produced reports (nor formally accepted) which purport that the organizations have the ability to calculate the nuclear control characteristics of hafnium and perform calculations that indicate hafnium, if properly designed and constructed, will privide essentially the same control characteristics as Ag-In-Cd, the control material presently used in Westinghouse reactors.
Verification of the above claims is nonesistent in the FASR and SER.
The density of the proposed hafnium neutron absorber is 0.467 lb/in3 versus 0.367 lb/in3 for the Ag alloy design (a 27% increase). Verification that the mainium control rods will operate satisfactorily under combined Loss of-Coolant Accidents plus Seismic Conditions as required by regulation is nonexistent in the FSAR and SER.
Verification that the proposed control rods will operate satisfactorily under vibration conditions, e.g. water hammer and /or load following conditions is nonexistent in the FSAR/SER.
a i
The only technical basis for use of a single failure criteria is the assumption
[
that each and every safety related function in Comanche Peak has a high probability of functioning properly so that if any one function fails all other safety functions can be counted on to prevent a serious accident. Verification that the full length I
jack control rod drive mechanisms will operate properly, especially ur. der load I
.h i
i s.
following conditions, with sufficiently high probability while using more dense hafnium control absorber material is also nonexistent in the FSAR /SER.
In each instance those topical reports referenced in the FSAR which purport to establish a basis for safe operation concerning the above are deficient in that the anticipated "as-built" condition is not verified.
II.
DELETION OF BORON INJECTION TANK According to Westinghouse literature, the boron injection tank is a primary component of one of two subsystems of the high head active safety injection system used in Westinghouse PWR's for the purpose of injection of highly borated water into the reactor collant system for small breaks which would result in slow blowdown and depressurization. The Applicant has deleted the boren injection tank at Comanche Peak.
This action has deleted the capability of injecting 15,000 ppm boric acid through dedicated piping into each of the four cold legs in relatively close proximity to the reactor vessel with the use of dedicated 1785 psi charging pumps. The sale remaining sybsystem (Chemical and Volume Control System) does not possess any of these attributes to the same extent as noted.
4 Justification for this action is that the only place the Applicant takes credit for this subsystem is in Chapter'15 of the FSAR during the analysis of large break accidents to satisfy the Design Basis Accident criteria. The justification is specious because the subsystem was designed to mitigate small break accidents.
3" Systems Summary of a Westinghouse Pressurized Water Reactor Nuclear-Power Plant", Westinghouse 4" Summary of Meeting on Comanche Peak Design Change and Responses to RSS Questions". Burwell, S.B., May 26, 1981
As a result of TMI-2, additional requirements have been imposed which include analyses of small-break loss-of-coolant accidents and of transients and accidents that involve postulated multiple failures, consequential failures, and operator errors, which if unmitigated could lead to inadequate core cooling. It is also required that the analyses be carried sufficiently into the event to identify all significant thermal / hydraulic /neutronic phenomena and shall address possibic failures and operator errors during the long-term cooling process (I.C. I).
The application for an operating license is deficient in that sufficient evidence has not been introduced to determine to what degree the boron injection tank deletion will reduce the margin of conservatism in the above analyses. (A rigorous overall conservatism has not been demonstrated, item 1.b, and in s.ne absence of this determination a marginal approach is the only technique available).
III.
LOFT AND SEMISCALE TESTS Requirement II.K. 3. 30 states that small-break loss-of-coolant accident analysis methods... shall be revised and provided that account for experimental data, including data from the LOFT and Semiscale test facilities.
A number of such tests indicate that the precise conditions experienced is the tests differed somewhat from those predicted.
There has been no basis submitted to justify why the required have not been accomplished.
IV.
OMISSION OF SAFETY EVALUATION REPORTS FOR TOPICAL REPORTS / COMPUTER CODES The NRC Staff and Westinghouse have elected to devote their resources to endeavors associated with Reduction In Conservatism (part Ib above). In the y.
e e
meantime, open issues concerning topical reports / computer codes languish.
Topical reports / computer codes dating back to 1971 which have been referred in the Comanche Peak FSAR have still not managed to make their way through the formal NRC review procedure and the associated NRC Staff Safety Evaluation reports on these generic reports are still not written. (The NRC Staff was unable to produce any documentatio6 at all in response to CFUR's Interrogatories).
If there are no substantial issues, these reports would have been formally accepted.
Nevertheless, construction of Comanche Peak has proceeded based, in part, on the conclusions of these reports. Were the reports found to have errors at this time, the Applicant could well be required to rebuild the plant - at least in part.
Now, at the eleventh hour, the NRC Staff reviewing the Comanche Peak submittal would have us believe that these reports are acceptable without the need a
for undergoing formal review and without the need for issuance of an NRC Staff Safety Evaluation Report. In short, without any basis.
The Comanche Peak Safety Evaluation Report is deficient in that there is no Generic Safety Evaluation Report on which to base conclusions regarding acceptability or non-acceptability of generic topical reports referenced in the Comanche Peak FSAR.
If the NRC Staff reviewing the Comanche Peak application wish to establish a basis for acceptance, they are required to issue a complete separate Safety Evaluation Report on each of the referenced Topical Reports or to include a complete plant specific Safety Evaluation Report concerning each Topical Report in the Comanche Peak Safety Evaluation Report..
~
}
The following list contains those topical reports listed in Section 1.6 of the Comanche Peak FSAR not related to Unresolved Safety Issues which have not been formally accepted, see also Attachments One and Three to "NRC Staff Answers to CFUR Seventh Set of Interrogatories", 2/3/82 and NUREG-0390, Vol. 5, No. 1, 1/20/82):
1.
"An Evaluation of Solid State Logic Reactor Protection in Anticipated Transients," WCAP-7706-L (Proprietary) and WCAP-7706 (Non-Proprietary), February 1973.
2.
" Overpressure Protection for Westinghouse Pressurized Water Reactors," WCAP-7769, Revision 1, June 1972.
3.
"LOFTR AN Code Description," WCAP-7907, June 1972.
4.
"FACTR AN - A FORTRAN-IV Code for Thermal Transients in a UO Fuel Rod," WCAP-7908, June 1972.
2 5.
" MARVEL, A Digital Computer Code for Transient Analysis of a Multiloop FWR System," WCAP-7909, June 1972.
6.
" Fuel Rod Bowing," WCAP-8691 (Proprietary) and WCAP-8692 (Non-Proprietat y), December 1975.
7.
" Properties of Fuel and Core Component Materials, WCAP-9179 (Proprietary), September 1977.
8.
" Evaluation of Nuclear Hot Channel Factor Uncertainties."
WCAP-7308-L (Proprietary) 'and WCAP-7810 (Non-Proprietary), December 1971.
9.
"Scismic Vibration Testing with Sine Beats," WCAP-7558, October 1971.
'10.
"Incore Power Distribution Determination in Westinghouse Pressurized Water Reactors. " WCAP-8498, July 1975.
11.
Failure Mode and Effects Analysis (F.EA) of the Engineered Safeguard Features Actuation System," WCAP-8584 (Proprietary) and WCAP-8760 (Non-Proprietary), Revision 1. February 1980.
-u-
12.
"Ger. eval Method of Developing Multi-frequency Bioxial Test Inputs for Distables," WCAP-8624 (Proprietary) September 1975 and WCAP-8695 (Non-Proprietary) August 1975.
13.
" Mass and Energy Releases Following a Steam Line Rupture,"
WCAP-8822 (Proprietary) and WCAP-8860 (Non-Proprietary), September 1976.
14.
" Bench Mark Problem Solutions Employed for Verification of WECAN Computer Program." WCAP-8929, June 1977.
15.
" Failure Mode and Effects Analysis (FMEA) of the Solid State Full Length Rod Control System." WCAP-6976, September 1977.
16.
" Westinghouse Emergency Core Cooling System Evaluation Model for Analyzing Large LOCA's During Operation with One Loop Out of Service for Plants Without Loop Isolation Values," WCAP-9166, February 1978.
17.
" Reactor Vessel Head Drop Analyses." WCAP-9198, January 1978.
Those topical reports / computer codes related to the TMI Action Plan also must meet these requirements. A partial list of such reports follows:
1.
"Nureg-0578, 2.1. 9.c, Transient and Accident Analysis",
2.
" Inadequate Core Cooling Studies of Scenarios with Feedwater Available, Using the NOTRUMP Computer Code", WCAP-9753.
3.
" Loss of Feedwater Induced Loss of Coolant Accident Analysis Report", WCAP-9744.
V.
HUMAN FACTORS The NRC Staff deals directly with Westinghouse in its determination of the safety analysis of generic reports. Many areas of concern such as the capability of a particular code to accurately predict a series of possible events are addressed l
l
)
in these dealings. A certain level of confidence is eventually established but the level of complexity of the possible physical processes are such that this level of confidence cannot be quantized with the use of current techniques. Yet, the NRC review of the Comanche Peak application makes no attmept to evaluate whether or not the Applicant is aware of the many uncertainties involved. The Applicant simply contracts with the NSSS supplier to originate all those portions of the FSAR pertaining to the NSSS. The supplier in turn references Topical Reports in whcih the Applicant has not been an active participant and, indeed, which the Applicant has not even totally read.
Now, as a result of TMI, a more thorough analysis is being required to take into account multiple and consequential failures and possible operator errors in both the initial and long term cooling process. But once again, the primary participants in this additional effort is the supplier, Westinghouse, and the NRC Staff. One of the expected results of this process is the determination of improved procedures. The Applicant will hire operators and evidently take action to insure that the operators act in consonance with those criteria agreed upon by Westinghouse and the NRC Staff.
One can only conclude that the Applicants personnel are supposed to act similar to a spinal cord whose brain resides in Washington and Westinghouse with an ill-defined nervous system connecting the two. So long as expected operating conditions are encountered, this arrangement is to prevail with absolutely no degree of innovation. But in the event that unexpected operating conditions exist, the Applicants personnel, with incomplete knowledge of all uncertainties, are expected to instantly become super-innovators to avert a serious accident.
However, when individuals exist in an atmosphere of non-innovation for long periods of time the expectation that these individuals will provide the desired..
super-innovation on instant demand is not very high. The same would apply to an individual paid by Westinghouse who spends 30 or 40 years of their working life in a trailer at Comanche Peak.
4 These human factors should be explicitly addressed in the Application for an Operating License, however no specific evidence that they are being appropriately addressed is available.
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EXHIBIT 2 CONTENTION SEVEN STATE.\\!ENT Contention Sev en provides:
" Applicants have failed to adequately evaluate whether the rock "overbreak" and subsequent fissure repair using concrete grant have impaired the ability of category I structures to withstand seismic disturbances."
CFUR's overall concern on this contnetion was whether or not the rock overbreak and subsequent grouting repairs affected the static and dynamic properties of the foundation for the plant and the plants ability to withstand seismic disturbances.
liowever during our investigation of this concern it was brought to our attention that the base for the mudmat for the foundation slab underlying Containment Unit One and/or the foundation of the safeguard's building may have been constructed improperly.
Namely one C. A. Thetford, a construction worker employed by Brown &
Root, was working on said mudmat and/or safeguard's building during the time period involved, stated to CFUR as follows:
The excavation for Unit One and/or safeguard's building was made too deep. Prior to pouring the concrete for the mud mat boulders which had previously been excavated from said hole were simply thrown loosely back into the hole and then the concrete was poured over them. Nothing was done to com; :.d said boulders and no grouting was done to said surface of the bottom of the hole in order to create a fl.t smooth surface to pour the mudmat on. The mudmat cencrete was poured in one continuous pour beginning on a Friday at 1:00 p.m. and continuing through to Sunday at 6: 00 p.m.
Said concrete mudmat pour constituted 6620 yards of continuously poured concrete. It was expressly stated by the witness that this concrete pour was conducted on the weekend in said fashion so that any deviations from regulations would not be detected.
4
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$y UNITED STATES OF AMERICA cW s NUCLEAR REGULATORY COMMISSION Ip i.
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
'lE D #
)
TEXAS UTILITIES GENERATING
)
Docket Nos. 50-445 COMPANY, et al.
)
50-446
)
(Comanche Peak Steam Electric
)
Station, Units 1 and 2
)
( Application for
)
Operating Licenses)
CERTIFICATE OF SERVICE I hereby certify that copies of the foregoing instrument in the above captioned matter were served upon the followigg persons gh;aw/
deposit in the United States mail, first class postage prepaid this'Q 3 day of, c-1982.
Marshall E. Miller, Esq.
Chairman, Atomic Safety &
Chairman, Atomic Safety and Licensing Appeal Panel I'icensing B ard U.S. Nuclear Regulatory Commission U.S. nuclear Regulatory Commission Washington, D.C. 20555 Washington, D.C. 20555 David J. Preister, Esq.
Dr. Kenneth A. McCollom Assistant Attorney General,,
Dean, Division of Engineering, Envir nmental Protection Division Architecture & Technology
.O. Box 12548 Capitol Station Oklahoma State University Austin, H 78711 Stillwater, Oklahoma 74074 J. Marshall Gilmore, Esq.
Dr. Richard Cole, Member 1060 W. Pipeline Atomic Safety & Licensing Board Hurst, n 76053 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Mrs. Juanita Elb.s President, CAST.
Chairman, Atomic Safety & Licensing 14 6 South k Street i
Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 7,ir. Chase R. Stephens Docketing & Service Branch Marjorie Ulman Rothschild, Eso.
U.S. Nuclear Regulatory Commission i
Office of the Executive Leg ' Director Waslu,ngton, D.C. 20555 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Richard E. Fouke Nicholas S. Reynolds Esq.
rlin gte,T h10 Debevoise & Liberman 1200 17th St., N.W.
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Washington, D.C.
20036 L
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J. MARSilALL GILMORE
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