ML033070233
| ML033070233 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 10/29/2003 |
| From: | Lochbaum D Union of Concerned Scientists |
| To: | Brian Benney Office of Nuclear Reactor Regulation |
| References | |
| 2.206, G20030545 | |
| Download: ML033070233 (22) | |
Text
Union of Concerned Scientists t *P. na' U-lUMim elveeA:e6lePee Citizens and Sientsts for Environmentsl Solutions DEDH DEDM AO October 29, 2003 DEDR Q
Mr. Brian J. Benney, Petition Manager OGC D Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission S;iealf.Q Washington, DC 20555-0001
SUBJECT:
INDIAN POINT ENERGY CENTER - SUPPLEMENT TO PETITION PURSUANT TO 10 CFR 2.206 - PWR CONTAINMENT SUMP FAILURE
Dear Mr. Benney:
By letter dated September 8, 2003, Riverkeeper, Inc. and the Union of Concerned Scientists petitioned the NRC to take enforcement action against the owner of the Indian Point Energy Center. Enclosed is a report prepared by UCS regarding the safety matter at the heart of our petition - the PWR containment sump problem. We are formally supplementing our petition to add the fact that Indian Point's owner is not complying with 10 CFR 50.46 regarding analysis and reporting of emergency core cooling system performance. This information has bearing on both the requested action already tossed out by the NRC -
namely, immediate shut down of both reactors pending correction of the safety problem - and the alternative requested action still being pondered by the agency.
As the report documents, the NRC took enforcement action in January 1996 when reasonable doubt (not absolute proof) was raised regarding Maine Yankee's conformance with 10 CFR 50.46. The NRC has not taken comparable action now when even stronger doubt was raised regarding Indian Point's conformance with 10 CFR 50.46.
For your information, by separate letter this date I am asking the NRC's Office of the Inspector General to investigate why the NRC failed to enforce 10 CFR 50.46 with respect to the PWR containment sump problem.
Sincerely, David Lochbaum Nuclear Safety Engineer Union of Concerned Scientists 1707 H Street NW, Suite 600 Washington, DC 20006 (202) 223-6133 (202) 223-6162, fax
Enclosure:
UCS report dated October 29, 2003, titled, "REGULATORY MALPRACTICE: NRC's "Handling" of the PWR Containment Sump Problem."
Washington Office: 1707 H Street NW Suite 600
- Washington DC 20006-3919
- 202-223-6133
- FAX: 202-223-6162 Cambridge Headquarters: Two Brattle Square
- Cambridge MA 02238-9105
- 617-547-5552
- FAX: 617-864-9405 California Office: 2397 Shattuck Avenue Suite 203
- Berkeley CA 94704-1567
- 510-843-1872
- FAX: 510-843-3785 Tr 4(OcV -405oI &016b5" 1o-
Union of Concerned Scientists Citizens and Scientists for Environmental Solutions REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem David Lochbaum, Nuclear Safety Engineer The Energy Reorganization Act of 1974 created the U.S. Nuclear Regulatory Commission (NRC). This Act of Congress gave the NRC the important, and exclusive, task of protecting the public from the radiological consequences from nuclear power. No other entity at the local, state, or federal level shares this responsibility with the NRC. In fact, all other entities are legally precluded from encroaching on the NRC's regulatory turf. Thus, there is no backup at any level to protect the public when the NRC does its job poorly.
This report exposes regulatory malpractice by the NRC regarding the pressurized water reactor (PWR) containment sump problem. This problem afflicts 68 of the 103 nuclear reactors operating in the United States and makes it much more likely that one of these reactors experiences the ultimate disaster -
meltdown with containment failure. This problem has been faced and fixed at literally dozens of PWRs around the world, but only one US PWR has been fixed to date. This report clearly documents that the NRC neither enforces its own regulations adopted to protect the public nor uses the legal means at its disposal to revise regulations it no longer cares to enforce. This report details how the NRC selectively uses risk arguments to enforce or dismiss regulations it sees as convenient. Additionally, this report examines the history of NRC's actions - and inactions - on this safety issue back to the 1970s.
The NRC's regulations are its strength: in general, these regulations are supposed to limit the risk from nuclear reactors to an acceptable level. The NRC's weakness is its inability to consistently enforce its regulations: like the assembly instructions for a bicycle or the directions for medication, regulations are only effective when they are followed. By failing to consistently enforcing its regulations, the NRC is not regulating the risk from nuclear power reactor operations to an acceptable level. The risk is, in fact, unacceptably high by the NRC's own standards.
The purpose of this report is not to chronicle yet another abysmal regulatory failure by the NRC. It certainly achieves that end, but that was not the objective. After all, it could be easily lost in the vast archives of similar reports by UCS, the General Accounting Office, the NRC's Office of the Inspector General, and many others. No, this report seeks to provide the United States Congress with compelling evidence that the NRC is shirking its unique responsibility of protecting public health and safety. It is our hope that the US Congress, which created the NRC with its Energy Reorganization Act, will expeditiously undertake the intensive series of hearings needed to reform the NRC so that it consistently enforces its safety regulations. In other words, we hope the US Congress will do now what it would do in its post-mortem inquiries into a tragic PWR accident - just skipping the part where thousands of Americans get harmed and a large region of our country gets ruined for decades.
Washington Office: 1707 H Street NW Suite 600 Washington DC 20006-3919 202-223-6133
- FAX: 202-223-6162 Cambridge Headquarters: Two Brattle Square
- Cambridge MA 02238-9105
- 617-547-5552
- FAX: 617-864-9405 California Office: 2397 Shattuck Avenue Suite 203
- Berkeley CA 94704-1567
- 510-843-1872
- FAX: 510-843-3785
REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem NRC's Failure to Enforce Regulations and Assure Public Safety Much has been discussed and written about terms like "adequate protection" and "reasonable assurance" as they apply to the regulation of nuclear power plant safety levels. The NRC's Atomic Safety and Licensing Appeal Board (ASLAB) best articulated these concepts, in our view, in this ruling:
As a general rule, the Commission 's regulations preclude a challenge to applicable regulations in an individual licensing proceeding. ... Generally, then, an intervener cannot validly argue on safety grounds that a reactor which meets applicable standards should not be licensed. By the same token, neither the applicant nor the staff should be pennitted to challenge applicable regulations,either directly or indirectly. Those parties should not generally be permitted to seek or justify the licensing of a reactor which does not comply with applicable standards. Nor can they avoid compliance by arguing that, although an applicable regulation is not met, tile public health and safety vill still be protected. For, once a regulation is adopted, the standards it embodies represent the Commission'sdefinition of what is requiredto protect the public health and safety. [emphasis added]
In short, in order for a facility to be licensed to operate, the applicant must establish that the facility complies with all applicable regulations.If the facility does not comply, or if there has been no showing that it does comply, it may not be licensed. [emphasis added]
The United States General Accounting Office, among others, acknowledged this ruling:
NRC's rules, regulations,and general design criteria (collectively referred to as regulations) are contained in Title 10, Chapter 1, Code of FederalRegulations. NRC regulationsareformal legal requirementsthat utilities must meet to constrictand operate theirplants. 2 [emphasis added]
As we understand this ruling, interveners - such as Riverkeeper and ourselves in the recently submitted 2.206 petition on the containment sumps at the Indian Point Energy Center - cannot argue that a reactor meeting all applicable regulations is unsafe, or more legalistically, that a reactor meeting all applicable regulations lacks "reasonable assurance" that the public has "adequate protection." Implicit in the ASLAB's ruling is the obligation on interveners to challenge a regulation they consider insufficiently safe while it is being promulgated. For once the NRC adopts a regulation, it defines "adequate protection" and compliance with it provides "reasonable assurance" of public safety. After its adoption, interveners can only challenge the adequacy of a regulation through the petition for rulemaking process.
As we understand this ruling, licensees - such as Entergy in the recently submitted Indian Point petition -
and the NRC cannot argue that a reactor failing to meet all applicable regulations is nevertheless safe.
Implicit in the ASLAB's ruling is the obligation on licensees to challenge a regulation considered unnecessarily burdensome while it is being promulgated. After its adoption, licensees can only challenge a regulation through the petition for rulemaking process. Likewise, the NRC cannot choose not to enforce a regulation after its adoption but must change a regulation deemed onerous through a rulemaking proceeding. No matter how well justified, non-compliance with an existing regulation fails to meet the "reasonable assurance" of "adequate protection" standard and violates the regulatory compact between the NRC and the public. When the NRC simply decides not to enforce a regulation, it violates the Administrative Procedures Act by essentially revising a regulatory standard without the legal necessities of public notice and public comment period.
Pre-eminent among many regulations applicable to the pressurized water reactor (PWR) containment sump issue is §50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear October 29, 2003 Page 2
REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem power reactors, of Title 10 in the Code of Federal Regulations. This regulation is relatively short and is provided in its entirety in Appendix A.
The Maine Yankee case illustrated how this regulation must be enforced. On December 4, 1995, the NRC received an anonymous letter sent to UCS and forwarded to the State of Maine. The author of the letter alleged that a computer code used to calculate the reactor system's response to postulated accidents had been deliberately manipulated to artificially inflate safety margins. The results from these allegedly
'fudged' analyses had been used to show that operation at an uprated power level of 2,700 Mwt met all requirements of 10 CFR 50.46. On January 3, 1996, the NRC issued an order to Maine Yankee's owner reducin the reactor's maximum output to 2,440 Mwt, the level established by the original operating license. The NRC had not substantiated the allegations before issuing the order, but the agency had concluded that the allegations constituted reasonable doubt about safe operation at 2,700 Mwt. Because the safety analyses for operation at 2,440 Mwt had not been done using the suspect computer code, the allegations did not raise reasonable doubt about operation up to that power level. Thus, the allegations removed the NRC's confidence that Maine Yankee satisfied 10 CFR 50.46 when operated at 2,700 Mwt but did not undermine their confidence that Maine Yankee complied with the regulation at power levels up to 2,440 Mwt. The NRC enforced 10 CFR 50.46 by ordering Maine Yankee to operate only under conditions where reasonable assurance existed.
This regulation, 10 CFR 50.46, is also applicable to the PWR containment sump issue as described by the NRC's Office of Nuclear Reactor Regulation:
The staff believes that there is sufficient new information and concerns raised relative to the potentialfor debris clogging in PWRs that this action plan has been updated to address PWR sump blockage concerns. As noted above, RESs [NRC's Office of Research] parametric evaluation demonstrated that sump blockage is a plausible concern for operating PWVRs. The results of the parametric evaluation form a credible technical basis for concluding that sump blockage is a potential generic concern for PWRs; however, the parametric evaluation is ill suited for making a determination that sump blockage will impede or prevent long-term recirculationat a specific plant. Therefore, it is not clear how significant a threat to PWR ECCS operation exists. The staff considers continued operation of PIVRs during the implementation of this action plan to be acceptable because the probabilityof the initiating event (ie., large break LOCA) is extremely low. More probable (although still low probability) LOCAs (small, intermediate) will generate small quantities of debris, require less ECCSflow, take more time to use ip the water inventory in the refueling water storage tank (RWST), and in some cases may not even require the use of recirculationfrom the ECCS sump because theflow through the break would be small enough that the operator will have sufficient time to safely shut the plant down. In addition, all PVRs have received approval by the staff for leak-before-break (LBB) credit on their largest RCS primary coolant piping. While LBB is not acceptable for demonstrating compliance with 10 CFR 50.46, it does demonstrate that LBB-qualified piping is of sufficient toughness that it will most likely leak (even under safe shutdown earthquake conditions) rather than rupture. ... And finally, the staff believes that continued operation of PVRs is also acceptable because of PIWR design features which may minimize potentialblockage of the ECCS sumps during a LOCA. The RES study on sump blockage attempted to capture many of the PWVR design features parametrically,however, it is not possible for a generic study of this nature to capture all the variationsin plant-specifcfeaturesthat could affect the potentialfor ECCS sump blockage (e.g., piping layouts, insulation location within containment, etc.). Therefore, evaluation on a plant-specific basis is necessary to determine the potentialforECCS sump clogging in each plant. 4 October 29, 2003 Page 3
REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem The NRC views the PWR containment sump issue as being a potential threat to Emergency Core Cooling System (ECCS) operation, but claims it lacks the plant-specific analyses needed to make determinations on individual reactors. Rather than make plant owners perform the plant-specific analyses, the NRC had Los Alamos perform pseudo-plant-specific analyses:'
As part of the GSI-191 study, RES's contractor, Los Alamos National Laboratory (LANL),
performed a generic risk assessment to determine how much core damage frequency (CDF) is changed by the findings of the parametric analysis. Utilizing initiating event frequencies that considerLBB credit consistent with NUREG/CR-5750, LANL calculatedan overall CDF of 3.3E-06 when debris clogging as afailure mechanism is not considered, and an overall CDF of 1.5E-04 when debris clogging is considered. ... The change in CDF is also dominated by the small and very small break LOCAs which are events where there are significant operator actions that can be taken to prevent core damage. 5 The two order of magnitude (i.e., factor of Slod 100) risk increase calculated by Los Alamos I 1.;r TaU LOCA 5-7 Ran MWOCA of Pamonwkt"ilE%
Itst0 uo 1v t
to ID Is ttga t
Lt0CA Potz.,U.
ICEg af a
1 for the operating fleet of 69 PWRs resulted UWkb_
W-r ft0.
totpe ,1 WY et 3vU 5
VM~L**0 k"*
%WY UkV
- Lb I
from their estimates of the likelihood that the "5404 V"1W
_N "LW V".7" by t 40 W A* 10h
" VY Lb containment sump screens would be clogged L "LA40 ftym 41 W .0 Lob U-We41 14040 42 V- L L tW4 by debris generated during a loss of coolant betU WY Ur *Vm MW _3 UA b to VL 41* oT OW W UL* 42 VI" b ey accident, thus preventing emergency pumps _4 y WL 4o A07Lb Doy i from supplying cooling water to the reactor. V.,*4LAW0, V"th 44 L VP- L tb 40,1~
W* -ah_ ?*V-T Lo bkdf Wy Los Alamos tabulated their estimates for each *ttbdr A W E ^ t A s of 69 PWRs for postulated small, medium, U"LA %L",4 o, 4I U , U,- I and large loss-of-coolant accidents (SLOCA, U." L~, Wke" S"1- J .- r 40 Lby I MLOCA, and LLOCA respectively). 6 About =VoA tV~y
" %-vL" m
L V"4LooF 45
£2 44 LA.V D I4 UN 41147 bV710 r LS 4040b eb w~
ar halftof the operating PWRs are "Very Likely" U-kft _
S tSf 5 L" Lard Vo t to experience blockage of the containment tuL"" ft vqV Dt" so y- A tykw "
sumps causing failure of the emergency core "erLA" PbW D~
FW L
-qtY%A Y-Y U d W 3 b W.-If VLA 0-
_ et L L.
cooling system pumps in event of either a ~ V-?
V~*~~*_ Wyur-ly
- b VL-i
- L; If* < _
Oa I small, medium, or large loss-of-coolant = GS V-Or VWobr4r V"v U te accident. bebEttIW" ft onh_ U-tat40 *0 b r e Vv7 5 47W*4 t U- _ j b0yI f £ I_ b The NRC's reticence in making plant owners perform the plant-specific analyses to confirm 7Va AD"A I 31oA It LCA or refute Los Alamos's results is inexplicable.
According to paragraph (a)(l)(i) of 10 CFR -2 50.46: L- .r.rq 55575400*57707551 707(fr7t.7 I' Oflhrers Except as provided in paragraph(a)(1)(i) of this section, the evaluation model must include sufficient supportingjustification to show that the analyticaltechnique realistically describes the behavior of the reactorsystem during a loss-of-coolant accident. Comparisons to applicable experimental data must be made and uncertaintiesin the analysis method and inputs must be identifled and assessed so that the uncertainty in the calculatedresults can be estimated. This uncertainty must be accountedfor, so that, when the calculated ECCS cooling performance is The studies are termed "pseudo-plant-specific" because they relied on plant-specific parameters like containment sump screen size, mesh size, emergency core cooling system flow rates, emergency core cooling system net positive suction head required and available, etc. but apparently lacked something preventing NRC from removing the "pseudo" prefix.
October 29, 2003 Page 4
REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem compared to te criteria set forth in paragraph (b) of this section, there is a high level of probabilitythat the criteriawould not be exceeded.
If nothing else, the Los Alamos results raise very serious, credible doubt about the realism in the evaluation models regarding the behavior of the reactor systems during loss-of-coolant accidents. After all, if it is "very likely" that the containment sump will clog, it is also "very likely" that the emergency core coolant system pumps will not be able to supply as much water to the reactor vessel as calculated (and more importantly, as needed to prevent reactor core damage). Paragraph (a)(2) of 10 CFR 50.46 provides the NRC with explicit power to ensure compliance with 10 CFR 50.46:
7he Director of Nuclear Reactor Regulation may impose restrictions on reactor operation if it is found that the evaluations of ECCS cooling performance submitted are not consistent with paragraphs(a)(I) (i) and (ii) of this section.
The Los Alamos results strongly suggested that the evaluations of ECCS cooling performance for most of the operating PWRs were NOT consistent with 10 CFR 50.46. The level of doubt raised by the Los Alamos studies surpasses by far the level of doubt raised by the Maine Yankee computer code allegations.
The NRC imposed restrictions on operation at Maine Yankee based on that "reasonable doubt." The NRC has, thus far, essentially ignored substantially greater "reasonable doubt" at the nation's operating PWVRs.
But the NRC's reluctance to enforce 10 CFR 50.46 should not have prevented plant owners from complying on their own. The public availability of the August 2001 Los Alamos study should have prompted each PWR owner to estimate the effect of the error in the evaluation model to satisfy 10 CFR 50.46 paragraph (3)(i) which states:
Each applicantfor or holder of an operating license or constructionpermit shall estimate the effect of any change to or errorin an acceptable evaluation model or in the applicationof such a model to determine if the change or error is signifcant. For this purpose, a significant change or erroris one which results in a calculatedpeakfuel cladding temperature different by more than 5OFfrom the temperature calculated for the limiting transient using the last acceptable model, or is a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperaturechanges is greaterthan 5001.
Los Alamos estimated that the core damage frequency increased by a factor of 100 for the average PWR due to the containment sump issue. The reactor core cannot be damaged unless the peak fuel cladding temperature difference exceeds 50'F, by bunches. 10 CFR 50.46 paragraph (3)(ii) specifies:
If the change or erroris significant, the applicantor licensee shallprovide this report within 30 days and include with the reporta proposedscheduleforproviding a reanalysis or taking other action as may be needed to show compliance with §50.46 requirements.
The NRC contends it lacks "absolute proof" that any of the 68 PWRs fails to comply with 10 CFR 50.46 and is unable to act absent that proof. But the NRC did not require "absolute proof" in the Maine Yankee case and doesn't need it now. Considering that compliance with 10 CFR 50.46 does not provide "absolute assurance" of public safety, only "reasonable assurance," it is ludicrous that the NRC would require "absolute" proof of a violation rather than "reasonable" proof. By doing so, the NRC affords its licensees greater protection than the public. The public only gets "reasonable assurance" against a nuclear disaster. The licensees can scoff at regulations unless the NRC finds "absolute" proof that it is happening.
October 29, 2003 Page 5
REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem Thus, owners of PWRs determined by Los Alamos to be "Very Likely" to experience containment sump blockage and consequential emergency core cooling system pump failure should have evaluated the impact on their 10 CFR 50.46 analyses and reported to the NRC, within 30 days, their proposed schedule for compensatory actions. The PWR owners are not complying with 10 CFR 50.46. The NRC is not enforcing compliance with 10 CFR 50.46. And the public is not getting "reasonable assurance" of "adequate protection."
Nearly two decades ago, the NRC revised the criteria it would use to determine if the containment sump design in the next PWVR to be built complied with 10 CFR 50.46:
The current 50% screen blockage assumption4 identified in Regulatory Guide (RG) 1.82, "Sumps for Emergency Core Cooling and Containment Spray Systems, " should be replaced with a more comprehensive requirement to assess debris effects on a plant-specific basis. The 50% screen blockage assumption does not require a plant-specific evaluation of the debris-blockage potential and usually will result in a non-conservative analysisfor screen blockage effects.
The staff has revised Regulatory Guide (RG) 1.82, Revision 0, "Sumnps for Emergency Core Cooling and Containment Spray Systems" and the Standard Review Plan Section 6.2.2, "Containment Heat Removal Systems" based on the above technical findings....The revised guidance will be used on Construction Permit Applications; Preliminary Design Approval (PDA) applications,and applicationsfor licenses to manufacture that are docketed after six (6) months following issuance of RG 1.82, Revision 1, and Final Design Approval (FDA) applications, for standardized designs which are intended for referencing in future Construction Permit Applications, that have not received approval at six (6) months following issuance of the RG 1.82, Revision 1. 7 leemphasis added]
Assuming it follows its own revised guidance, the NRC will never again license a pressurized water reactor to operate without a plant-specific evaluation of the debris-blockage potential for the containment sump screens. In revising Regulatory Guide 1.82, the NRC implicitly conceded that the process it had used to show compliance with 10 CFR 50.46 for the nation's operating PWRs was faulted. Yet the NRC failed then, and has failed since then, to enforce another of its regulations, 10 CFR 50.100, Revocation, suspension, modification of licenses and construction permits for cause, of Title 10 in the Code of FederalRegulations which states:
A license or constructionpermit may be revoked, suspended, or modifed, in whole or in part, for any materialfalse statement in the applicationfor license or in the supplemental or other statement offact requiredof the applicant; or because of conditions revealed by the application for license or statement offact or any report, record, inspection, or other means, iehklichmdd warrant the Commission to refuse to grant a license on an original application (other than those relating to §§50.51, 50.42(a), and 50.43(b) of this part); or for failure to construct or operate a facility in accordance with the terms of the constriction permit or license, provided that failure to make timely completion of the proposed construction or alteration of a facility under a constructionpermit shall be governed by the provisions of §50.55(b); orfor violation of orfailtre to observe, any of the terms and provisions of the act, regulations, license, pennit, or orderof the Commission. [emphasis added]
Basically, this assumption had plant owners calculate how much water their emergency pumps could get from flow through the unblocked half of the containment sump screens. There was no effort by plant owners or NRC to verify that the containment sump screens would not get blocked more than 50 percent.
October 29, 2003 Page 6
REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem By its own admission, the NRC would refuse to grant an operating license to any future PWR applicant who had submitted a design based on the old 50 percent containment sump screen blockage assumption in Regulatory Guide 1.82 Revision 0. But the NRC opted not to revoke, suspend, or modify the operating license for any of the PWRs it had already licensed using that flawed basis. Why not? Robert Ripley would not believe this one:
However, the staffs regulatory analysis NUREG-0869, Revisio 1 "USI A43 Regulatory Analysis") evaluated (1) containment designs and their survivability should loss of recirculationoccur, (2) alternate means to remove decay heat, (3) release consequences (which were based on pipe break probabilities which did not incorporateinsights gainedfrom recent pipe fracture mechanics analyses), and (4) cost estimates for backfits considered (i.e.,
reinsulating). This regulatoryanalysis did not support a generic backfit action and resulted in the decision that this revised regulatory guidance will not be applied to any plant now licensed to operate or that is under construction. 8 [emphasis added]
The NRC relied on pseudo-plant-specific analyses in deciding NOT to fix the PWR containment sump problem in 1985. Apparently, the NRC viewed the results from such analyses good enough then to justify taking no action. It's curious that the NRC now views the results from pseudo-plant-specific analyses performed by Los Alamos showing that the PWR containment sumps must be fixed as not being good enough to justify taking immediate action. Pseudo-plant-specific analyses have the same merit in 2002 as they did back in 1985. Yet the NRC will reject analytical results showing there is a safety problem and embrace analytical results, of equivalent quality, indicating the lack of a problem. This is a shameful practice on the NRC's part that needs to be eliminated.
The pseudo-plant-specific analyses performed by Los Alamos concluded that the containment sump problem posed a serious safety hazard at the majority of the nation's operating PWRs. The only genuine plant-specific analysis, to our knowledge, was performed for the Davis-Besse nuclear plant:
The [Davis-Besse] licensee reported in LER 346/2002-005 deficiencies with the containment emergency sump to perform its function under certain accident scenariosdue to clogging of the emergency core cooling and containment spray systems' snp screen by fibrous materials, unqualified coatings, and various other debris.
The licensee performed comprehensive and extensive corrective actions. The actions included installation of a larger emergency sump strainer,field walk-downs of potential debris which could potentially clog the emergency sump, and evaluations on potential debris left in containment. Additionally the licensee has analyzed the emergency core cooling system and containment spray system recirculationfunctions with respect to the potentially adverse post-accident debris blockage effects to confirm compliance with 10 CFR 50.46(b)(5) and all other existing applicableregulatoryrequirements.
The NRC conducted inspections and evaluations to verify that the containment sump deficiencies had been adequately resolved. The NRC reviewed the licensee's corrective actions, including the design modification for the new sump, field installation, and compliance with regulatory requirements. The NRC inspections are documented in Inspection Report 50/346 06 and in this Inspection Report.
In September 2, 2003, the Davis-Besse Oversight Panelmet to discuss this issue and concluded that Restart Checklist Item 2.c.] is closed. 9 [emphasis added]
October 29, 2003 Page 7
REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem These recent events at Davis-Besse confirm what Los Alamos suspected - that the PWR containment sump is vulnerable to clogging with consequential impairment of the reactor core cooling and containment cooling systems. In fact, Los Alamos had determined Davis-Besse to be one of the PWRs least susceptible to the containment sump problem, but the problem at Davis-Besse was very real and had serious safety consequences - the NRC issued a YELLOW finding" to FirstEnergy for Davis-Besse containment sump issue.' 0 The events at Davis-Besse demonstrate beyond any doubt that plant owners can analyze debris generation, debris transport, debris loading on containment sump screens, and resulting impact on safety system performance and that the NRC can review that analysis and determine if it complies with 10 CFR 50.46. Since it is undeniable that the plant-specific analyses can be performed and reviewed today, there are no legal grounds for continuing to postpone these analyses. This work can be done - the NRC is meekly tolerating it not being done.
The NRC is simply not enforcing 10 CFR 50.46 with regard to the PWR containment sump issue.
Enforcing compliance with this federal regulation is the only legal way for the NRC to properly discharge its responsibility to protect public heath and safety:
For, once a regulation is adopted, the standards it embodies represent the Commission's definition of what is required to protect the public health and safety. IX[emphasis added]
It is ironic that the NRC is guilty of the same root cause that it assigned to Northeast Utilities' failings at the Millstone Nuclear Power Station:
While there is a strong emphasis on safety as a stated objective, the organization does not consistently recognize or emphasize the collective set of administrative (e.g., the proposed Determine Course ofAction (DCA) concept) and technicalprocesses (e.g., Setpoint Control)that demonstrate and assure that objective is met. 12 It is unbelievable that the NRC is also guilty of the same sin that its own Special Inquiry Group (also known as the Rogovin committee) identified as a significant contributing factor to the Three Mile Island accident:
[Categorization of an issue as generic typically delays its resolution. Because issues are regarded on a generic basis and are not regarded as an impediment to individual plant licensing, little incentive exists for their resolution. 13 The NRC should be ashamed that it cannot find incentive for resolving generic safety hazards, such as the PWR containment sump problem, in assuring adequate protection of millions of Americans. Instead, the NRC gambles, once again, with the lives of Americans:
The Davis-Besse plant, located in Oak Harbor,Ohio, received an NRC operating license in 1977.
In 1979 NRC inspected the plant and recommended that the utility install a third auxiliary feedwater pump to correct a design deficiency that NRC concluded could contribute to a core melt during an accident.... However, NRC alloved the utility time to analyze alternativessuch as upgradingprocedures and control systems before taking te required corrective action. In 1984, S years later, the utility agreed to install the thirdpump by late 1985. However, before the pump itas installed, the plant experienced a series of equipment failures and operator errors in June 1985 such that (1) one mainfeedivater pump became inoperable, (2) the utility could not activate Beginning in April 2000, the NRC employs a four-color rating scheme to safety findings - GREEN, WHITE, YELLOW, and RED in order of increasing significance.
October 29, 2003 Page 8
REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem anotherfeedvater pump, and (3) the auxiliaryfeedvater system became inoperable. In October 1985 and May 1986 hearings before the Subcommittee on Energy Conservation and Power, House Committee on Energy and Commerce, NRC admitted that although the equipment probleinsposed an undue risk to public health and safety, the agency waited too long to require the utility to install the third pump. NRC also admitted that its inspection and enforcement programfailed to identify the potential for the extensive equipment failures that subsequently occurred, even though the plant's operating perfonnance had declined since 1982. 14 [emphasis added]
The NRC consistently asserts public health and safety is its top priority, consistently promulgates regulations to provide adequate protection of public health and safety, and inconsistently enforces those regulations. The NRC allows nuclear reactors to routinely operate in violation of the regulations specifically adopted to provide "reasonable assurance" the public has "adequately protection." Thus, the NRC is failing to demonstrate and assure that its objective is being met. It is waiting too long to resolve the PWR containment sump issue. It is exposing millions of Americans to higher risk than is necessary.
Improper Reliance on Probability Arguments Following receipt of a 2.206 petition on the containment sump problem at the Indian Point Energy Center,' 5 NRC staffers have been quoted by the media asserting that the low likelihood of an accident requiring the containment sumps to "save the day" justifies the long-term resolution plan for the issue. At least one NRC staffer was quoted as saying that there's never been a prior event at any PWR where the containment sump was challenged. UCS fails to see the relevance of these NRC statements. Our version of 10 CFR Part 50 doesn't contain footnotes or other caveats explaining that the regulations are only applicable after the first accident. Best we can tell, the regulations are intended to protect the American public from even the first accident.
Earlier this year, the NRC issued an order to every licensee (i.e., all the PWRs and all the BWRs) mandating immediate actions be taken to upgrade security measures:
As part of the Commission's review of the security and safeguardsprogram, the Commission has assessed information providedby the intelligence coinimnity and determined that revisions to the DBT [design basis threat], as currently specifed in Title 10 of the Code of FederalRegulations,
§73.1(a), are required. The Commission has determined that the current threat environment requiresthat the enclosed Order be effective immediately. 16 [emphasis added]
No US nuclear plant has ever experienced an event where security was challenged by a terrorist. No publicly available risk assessment suggests, yet along demonstrates, that the terrorist threat is anywhere close to the core damage risks estimated by Los Alamos for the PWR containment sump issue. No publicly available threat assessment suggests that the probability of terrorist attack is nearly as high as the probability of a loss of coolant accident at a PWR. Yet the NRC took action on the security issue and inaction on the PWR containment sump problem.
The NRC is arbitrarily and capriciously using probability in its regulatory decision-making. When it received bad press, the NRC set aside the probability numbers and ordered plant owners to immediately fix security problems. Until it receives bad press, the NRC invokes the probability numbers to postpone fixes of the PWR containment sump problems. The NRC should be more concerned about public safety than its Neilsen ratings.
October 29, 2003 Page 9
REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem Containment Sump Issue History Generic Safety Issue No. 191 (GSI-191) is merely the latest incarnation of the NRC's process for NOT resolving the PWR containment sump problem that has plagued US nuclear plants for over two decades.
NRC Chairman Joseph Hendrie wrote to President Jimmy Carter on Valentine's Day 1979 about it:
Following a postulated loss-of-coolant accident, i.e., a break in the reactor coolant system piping, the waterflowing through the break would be collected in the emergency sutnp at the low point in the containment. This water would be recirculated through the reactor system by the emergency core cooling pumps to maintain core cooling. Tis water would also be circulated through the containment spray system to remove heat andfission productsfrom the containment.
Loss of the ability to draw water from the emergency sunp could disable the emergency core cooling and containment spray systems. The consequences of the resulting inability to cool the reactor core or the containment atmosphere could be melting of the core and/or breaking of the containment.
One postulated means of losing the ability to draw water from the emergency sumnp would be blockage by debris. A principal source of such debris could be the thennal insulation on the reactorcoolant system piping. In the event of a pipe break, the subsequent violent release of the highpressure water could rip off the insulation in the area of the break. This debris could then be swept into the sumps potentially causing damage. 17 At the time NRC apprised President Carter of a problem that could lead to reactor meltdown, dozens of PWRs were under construction in the United States including many of those operating today. Rather than ensure that each, or any, of these reactors corrected the known safety problem before it commenced operating, the NRC treated the problem as an Unresolved Safety Issue, specifically Issue A-43:
Unresolved Safety Issue (USI) A43 deals with safety concerns related to the availability of adequate recirculation cooling waterfollowing a loss-of-coolant accident (LOCA) when long-term recirculationcooling must be maintainedto prevent core melt. These safety concerns can be summarized by thefollowing question:
In the recirculation mode will the sump design (for pressurized water reactors,PWRs) or the residual heat removal (RHR) suction intakes (for boiling water reactors,BWRs) provide sufficient water to the RHR pumps, and will this water be sufficiently free of LOCA-generateddebris and airingestion so as not to impairpump performance, thereby providing adequate net positive suction head (NPSH) to the recirculationpumps?
Generic plant insulation surveys, development of debris estimation and transport methods, experimnents, and public comments received have shown that debris effects are dependent on the type and quantities of insulation employed, the layout of the containment, and the post-LOCA recirculation requirements.A single, or overall generic, solution is not possible. 18 [emphasis added]
Given that the NRC's efforts between 1979 and 1984 concluded that a generic solution was not possible, one wonders why the NRC opened Generic Safety Issue No. 191 in September 1996 to deal with the PWR containment sump problem. After all, using the generic safety issue process to resolve a safety hazard for which you have already determined "A single, or overall generic, solution is not possible" seems like a bureaucratic stall tactic. If so, it has worked exceedingly well. If not, the NRC's glacial pace on GSI-191 has reaffirmed their conclusions from the 1980s.
October 29, 2003 Page I10
REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem Before the NRC closed USI A-43, they researched the problem and outlined three possible solutions:
Industry impact as a result of the proposed actions will vary from plant to plant because of the variationsin insulation employed, actual sump design and operationalrequirements (e.g., NPSH margins must be maintainedfor the recirculation pumps). Thus the impacts can be categorized as:
(1) analysis only required (2) analysis plus minimal plant backfits required (e.g., enlargement of screen area to trap debris, reduction of recirculationflow ratesfrom current ratings,etc.)
(3) analysis plus replacement of large quantities of the problem insulation required 19
[emphasis added]
All of the recommended options involved at least a plant-specific analysis. The NRC also looked at the option of doing nothing, but found ample reason to rule it out:
Ignoring the implications of the results of the USI A-43 debris-blockage effects study with respect to OLs [operating licenses] and NTOLs near-term operating licenses] is not acceptable. Sump failure probabilities of 3x]06 to 104/Rx-yr have been calculated; the results are very strongly plant dependent. The uncertainties associated with types and quantities of insulation utilized (and given the fact that there have been unreportedchanges in the types of insulation used) warrantfollowup action. 20 [emphasis added]
Unless obfuscation and foot-dragging count, the NRC did not do any follow-up action. The NRC closed USI A43 in 1985 without requiring any operating PWR to fix, or even analyze, anything. 21 The NRC somehow accepted that which it had determined was "not acceptable." The do-nothing option, which had been specifically excised from the ballot, somehow won the election:
... the staff has concluded that no new requirements need be imposed on licensees and constructionpermit holders... 22 [emphasis added]
The NRC, which professes to have public safety as its foremost priority, fabricated Unresolved Safety Issues for the express purpose of allowing nuclear power reactors to start up without fixing known safety problems. Mr. J. Montgomery of the NRC staff explained this gaming system to the Commissioners:
One thing I wanted to explain tat one of the limitation tat we discussed here is a limitation applied to consideration of certain kinds of costs in dealing with issues that were left umresolved at the construction pennit stage. That is a result, as Jim has pointed out, of a rather arcirane legal system but that system was developed deliberately as a accommodation to the nuclear power industry. It was developed in order to get plants licensed without resolving significant safety questions as in the PRDC case. In effect, a kind of deal was struck very early in the industry. 23 [emphasis added]
Thus, the NRC intentionally and deliberately avoided protecting the public and instead concentrated on merely getting the nuclear power plants up and running. The NRC has admitted to placing production ahead of safety.
Accommodating the industry at the expense of the public did not stop when the reactors started. A decade later, NRC Chairman Ivan Selin expressed his disappointment over the ongoing accommodation game:
October 29, 2003 Page
REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem
... a related area in which the industry must do better is in anticipating generic problems and in solving them early. This need will become more acute as the universe of regulated reactors gets older and new generic aging issues emerge. ... When confronted with the problem, te industry's response was to deny its existence without investigation,forcing the NRC to spend much time and resources to prove the problem's existence.... Later, when the NRC was able to prove that its concern was valid, both of us found ourselves in a position where a safety issue had been known for several years, but corrective action had not yet been taken. ... When generic problems such as these are not promptly and fully addressed, both the NRC and the industry find themselves under justifiable criticism. Additionally, unnecessary financial and organizational resources are often required to deal effectively ith such long-festering problems.2 4 [emphasis added]
Chairman Selin was referring to the industry's response, or non-response, to the motor-operated valve problem (i.e., Generic Letter 89-10), but his words apply equally to the PWR containment sump problem and a host of other safety issues. Why was the NRC Chairman even directing his staff "to spend much time and resources to prove the problem 's existence"? As the NRC points out:
"...it is important to recognize that although the staffs review of an application is partially an "audit" reviewn the applicantfor a license is obligated to assure compliance with applicable regulatory requirements. It is the applicant who bears the burden of proof on the issue." 25
[emphasis added]
Why has NRC usurped the obligations from its licensees? Why has NRC stolen the burden of proof from its licensees? An NRC manager posed this question before the US House of Representatives:
My reaction was, "Who te hell is regulating who? " In my opinion, the NRC did not have to buckle to industrypressure. 26 [emphasis added]
Likewise, the NRC buckled to industry pressure and closed USI A-43 without taking any action to resolve the safety issue at any operating PWR. The NRC claimed it needed plant-specific analyses to define what corrective actions were needed at what reactors. Okay, fair enough. But how could the NRC dismiss the issue without having performed a single plant-specific analysis? If a plant-specific analysis was necessary to definitely demonstrate presence of a safety threat, wouldn't a plant-specific analysis also be required to definitively demonstrate absence of that safety threat? Of course, but that wouldn't accommodate the nuclear industry. UCS hereby accepts Chairman Selin's implied invitation to justifiably criticize the NRC for not promptly and fully addressing the PWR containment sump issue since its inception in 1979: the NRC's performance on this important safety matter is atrocious.
Following the Three Mile Island meltdown in 1979, several task forces were empanelled to study the accident and its causes. President Jimmy Carter appointed Dartmouth College President John G. Kemeny to chair one of the task forces. Kemeny's committee concluded:
It [the NRC] was clearly not part of the solution but a serious part of the problem. ... They had the lovely habit of giving some very difficult issues a special "generic" label, thus allowing these issues to sit on the shelf. 27 [emphasis added]
A student in school cannot hope to avoid missing questions on an examination by labeling them "generic" and thus unanswerable. A taxpayer in an Internal Revenue Service audit cannot hope to explain questionable deductions by labeling them "generic" donations and thus receipt-free. But nuclear plant October 29, 2003 Page 12
REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem owners know they can avoid fixing safety problems by labeling them "generic" and forcing NRC to play the accommodation game. By playing along, the NRC is aiding and abetting practices subjecting millions of Americans to unnecessary risk.
The NRC's irrational behavior is best explained by someone who served on the Nuclear Regulatory Commission at the time when the PWR containment sump problem was first buried. Commissioner James Asselstine bemoaned and protested the NRC's intentional and deliberate abandonment of its mission to protect public health and safety:
I can tink of no other instance in which a regulatoryagency has been so eager to stymie its own ability to carry out its responsibilities.Indeed, the adoption of this backfititng] rule is tie most compelling evidence to date of the Commission majority's open hostility to the regulatory mission of this agency.... By this step, the Commission is moving in the wrong direction - a direction that will likely result in further serious operating events, more accidents, and a lower level of safety than that achieved in many more forward-thinking countries in the world. 28
[emphasis added]
Lest Commissioner Asselstine's remarks are misconstrued as "sour grapes" whining from someone on the short end of a Commission vote, let's examine what "more forward-thinking countries in the world" have already DONE about the PWR containment sump problem: 29 Country Completed Actions from 1992 to 2002 Belgium 6-fold increase in containment sump screen areas completed at 2 of 7 operating PWRs Canada 13-fold increase in containment sump screen areas completed at 4 of 18 operating PWRs; schedules for similar modifications at 10 other PWRs Czech Republic 4-fold increase in containment sump strainer area completed at 4 of 6 operating PWRs Finland Approximate 10-fold increase in containment sump strainer area completed at 2 of 2 operating PWRs Hungary Approximate 50-fold increase in containment sump strainer area completed at 4 of 4 operating PWRs Japan Approximately 95% of the fibrous insulation replaced by non-fiber insulation at 23 of 23 operating PWRs Netherlands 50 percent increase in containment sump strainer area completed at I of I operating PWVR Russia Significantly larger containment sump strainers installed at 4 of 6 operating PWRs Slovak Republic Significantly larger containment sump strainers installed at 4 of 4 operating PWRs So, while the NRC accommodates, the nuclear regulatory bodies in forward-thinking countries fix! At least 25 PWRs outside the US have resolved the containment sump problem by increasing the physical size of their sump screens to make them less vulnerable to clogging. At least 23 PWRs outside the US have resolved the containment sump problem by replacing insulation inside containment with a type more resistant to becoming debris in event of an accident. Only I PWR in the US has resolved the containment sump problem. The remaining 68 PWRs in the US are hiding behind the NRC's shield, which is supposed to protect the public but instead is being misused to prevent the public from getting the "adequate protection" it so richly deserves.
October 29, 2003 Page 13
REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem Conclusion and Recommendations Millions of Americans are exposed to higher risk than necessary because the NRC is not enforcing federal safety regulations. The regulations are the NRC's only legal means for determining with reasonable assurance that the public is adequately protected from the radiological consequences of nuclear power reactor operation.
The United States Congress should undertake hearings and related actions necessary to:
The PWR containment sump problem is well-known: NRC has been examining it since the 1970s.
The PWR containment sump problem's solution is equally well-known: NRC already oversaw resolution of a very similar problem at the nation's boiling water reactors and the resolution of this specific problem at the PWR at the Davis-Besse nuclear plant in Ohio. Literally dozens of PWRs outside the United States have already resolved the problem.
The NRC must cease its accommodation game with the nuclear industry and join the rest of the planet in fixing the PWR containment sump problem.
- 2. Review close-outs of other generic safety issues to determine if there are any other improperly closures.
The PWR containment sump problem was initially handled by the NRC as Unresolved Safety Issue A-43, which they closed in 1985 without requiring any actions to be taken by plant owners.
It is abundantly clear that NRC erred in its close-out process on this safety issue; hence the resurrection of this problem as GSI-191 in September 1996. All other safety issues that were closed without requiring any actions to be taken must be reviewed to determine if the PWR containment sump problem was the NRC's only such mistake.
- 3. Reform the NRC so that it is consistently enforcing regulations.
Some years ago, the NRC realized that its inspection program focused on the adequacy of procedures developed for nuclear plants and slighted how well workers followed those procedures. The NRC revamped its inspection procedures to examine outcomes as well as processes. Likewise, the NRC could develop metrics on its own outcomes. For example, the recent Lessons Learned Task Force report prepared by the NRC identified numerous regulatory failures to ensure compliance with regulatory requirements at the Davis-Besse nuclear plant. Such efforts should be supplemented by formal evaluations during special inspections, augmented inspections, and incident investigations conducted by the NRC in response to plant events to determine if inconsistent enforcement contributed to the events' occurrence and/or significance.
But the point is that when the NRC does not assess its own enforcement effectiveness, it can never identify - and more importantly, never correct - any shortfalls.
- 4. Reform the NRC so that it is resolving safety issues expeditiously.
The NRC has been moving towards what it calls risk-informed regulation and has developed plant-specific risk assessment models. For each safety issue, the NRC can determine how many October 29, 2003 Page 14
REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem operating reactors it potentially effects and at least a bounding estimate for the safety significance of the problem. This information would help the NRC prioritize and manage the resolution of multiple safety issues. For example:
Emerging Issue Reactors Affected Potential Safety Overall Safety Impact (number) Impact (per reactor)## (per industry)
A 68 8 544 B 103 2 206 C 9 40 1 360 Such an approach would help the NRC manage the resolution of safety issues on both a macro and micro level. On the macro level, the approach would demonstrate that Emerging Issue B, while affecting the entire fleet of reactors, has less overall safety impact than either Issue B or Issue C, which only affect a subset of the fleet. On the other hand, the approach would reveal that taking three times as long to resolve Issue B than Issue A has adverse safety consequences. Thus, the approach would better enable NRC to apply resources where they yield the greatest safety benefit.
On the micro level, the approach would highlight, for example, that the nine reactors affected by Issue C have a potentially significant impairment that should be considered by the NRC when making regulatory decisions about these reactors. Increased awareness of such potential impairments make it less likely that the NRC would approve a measure increasing the chances of challenged the potentially impaired system at these reactors.
Over its history, the NRC has separated resolution of emerging or generic safety issues from oversight of operating reactors. Integration of safety issue resolution into reactor oversight would expedite resolution. Plant owners would have great incentives either to demonstrate that their reactors are not affected by the issue or to fix the problem so that it doesn't continue to complicate other operational activities.
The NRC could undertake these reforms without Congressional involvement. But it could have done so any time during the past three decades yet failed to do so. The Congress must get involved now to make these overdue reforms happen and happen soon.
- These unitless, arbitrary values are provided to illustrate the concept.
October 29, 2003 Page 15
Union of Concerned Scientists Citizens and Scientists for Environmental Solutions REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem APPENDIX A: 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors (a)(J)(i) Each boiling or pressurized light-water nuclear power reactor fueled vith uranium oxide pellets within cylindricalzircaloy or ZIRLO cladding* must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteriaset forth in paragraph(b) of this section. ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculatedfor a number of postulated loss-of-coolant accidents of different sizes, locations, and other propertiessufficient to provide assurancethat the most severe postulatedloss-of-coolant accidents are calculated. Except as provided in paragraph(a)(1)(i) of this section, the evaluation model must include sufficient supportingjustification to show that the analyticaltechnique realistically describes the behavior of the reactorsystem during a loss-of-coolant accident. Comparisons to applicable experimental data must be made and uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated results can be estimated. This uncertainty must be accountedfor, so that, when the calculatedECCS cooling performance is compared to the criteriaset forth in paragraph(b) of this section, there is a high level of probability that the criteria would not be exceeded. Appendix K, Part II Required Documentation, sets forth the documentation requirementsforeach evaluation model. [emphasis added]
(ii) Alternatively, an ECCS evaluation model may be developed in conformance with the required andacceptablefeatures of appendix K ECCS Evaluation Models.
(2) The Directorof NuclearReactorRegulation may impose restrictionson reactoroperation if it is found that the evaluations of ECCS cooling performance submitted are not consistent with paragraphs(a)(l) (i) and (ii) of this section. [emphasis added]
(3)(i) Each applicantfor or holder of an operatinglicense or constructionpermit shall estimate the effect of any change to or errorin an acceptable evaluationmodel or in the applicationof such a model to determine if the change or erroris significant. Forthis purpose, a significant change or error is one which results in a calculatedpeak fuel cladding temperature different by more than 5OFfrom the temperature calculatedfor the limiting transient using the last acceptable model, or is a cumulation of changes and errorssuch that the sum of the absolute magnitudes of the respective temperaturechanges is greaterthan 507F. [emphasis added]
Indian Point Units 2 and 3 are both pressurized light-water nuclear power reactors fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding.
REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem (ii) For each change to or error discovered in an acceptable evaluation model or in te application of such a model that affects the temperature calculation, the applicant or licensee shall report the nature of the change or error and its estimated effect on the limiting ECCS analysis to the Commission at least annually as specified in §50.4. If the change or error is significant, the applicant or licensee shall provide this report within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with §50.46 requirements. This schedule may be developed using an integratedscheduling system previously approvedfor the facility by the NRC. For thosefacilities not using an NRC approved integratedscheduling system, a schedule vill be established by the NRC staff within 60 days of receipt of the proposedschedule. Any change or error correction that results in a calculated ECCS performance that does not conform o the criteria set forth in paragraph (b) of this section is a reportable event as described in §§50.55(e), 50.72 and 50.73.
The affected applicantor licensee shall propose immediate steps to demonstrate compliance or bring plant design or operation into compliance with §50.46 requirements. [emphasis added]
(b)(1) Peak cladding temperature. The calculated maximum fiel element cladding temperature shall not exceed 2200° F.
(2) Maximum cladding oxidation. Te calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation. As used in this subparagraph total oxidation means the total thickness of cladding metal that would be locally converted to oxide if all the oxygen absorbed by and reacted with the cladding locally were converted to stoichiometric zirconium dioxide. If cladding rupture is calculatedto occur the inside surfaces of the cladding shall be included in the oxidation, beginning at the calculated time of rupture.
Cladding thickness before oxidation means the radial distance from inside to outside the cladding, after any calculated ntptre or swelling has occurred but before signiflcant oxidation.
IVhere the calculated conditions of transient pressure and temperature lead to a prediction of cladding svelling, with or without cladding rupture, the unoxidized cladding thickness shall be defined as the cladding cross-sectionalarea, taken at a horizontalplane at the elevation of the rupture, if it occurs, or at the elevation of the highest cladding temperature if no rupture is calculated to occur divided by the average circumference at that elevation. For nptured cladding the circumference does not include the rupture opening.
(3) Maximum hydrogen generation. The calculated total amount of hydrogen generatedfrom the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that svould be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
(4) Coolablegeometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling.
(5) Long-term cooling. After any calculated successful initial operation of the ECCS the calculatedcore temperatureshall be maintainedat an acceptably low value anddecay heat shall be removedfor the extended period of time required by the long-lived radioactivityremaining in the core.
(c) As used in this section: (1) Loss-of-coolant accidents (LOCA's) are hypothetical accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system, from breaks in pipes in the reactorcoolant pressure boundary up to and October 29, 2003 Page 2
REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem including a break equivalent in size to the double-ended ruptureof the largestpipe in the reactor coolant system.
(2) An evaluation model is the calculationalframework for evaluating the behavior of the reactorsystem during a postulated loss-of-coolant accident (LOCA). It includes one or more computer programs and all other information necessary for application of the calculational framework to a specific LOCA such as mathematical models used, assumptions included in the prokrams. procedure for treating the program input and output information, specification of those portions of analysis not included in computer programs, values ofparameters, and all other information necessary to specify the calculationalprocedure.[emphasis added]
(d) The requirements of this section are in addition to any other requirements applicableto ECCS setforth in this part. The criteriaset forth in paragraph(b), with cooling performancecalculated in accordance with an acceptable evaluation model, are in implementation of the general requirements with respect to ECCS coolingperformance design setforth in this part,including in particularCriterion35 of appendixA. 30 October 29, 2003 Page 3
REGULATORY MALPRACTICE: l NRC'S "Handling" of the PWR Containment Sump Problem l Sources:
1Atomic Safety and Licensing Appeal Board, Atomic Energy Commission, Memorandum and Order (ALAB-138),
Section IV, paragraph A.1, July 31, 1973.
2United States General Accounting Office, GAO/RCED-87-141, "Nuclear Regulation: Efforts to Ensure Nuclear Power Plant Safety Can Be Strengthened," page 10, August 1987.
3 Office of the Inspector General, Nuclear Regulatory Commission, Case No.97-03S, "Public's Concerns with NRC Report of Independent Safety Assessment (ISA) at Maine Yankee," January 26, 1998.
4Office of Nuclear Reactor Regulation, Nuclear Regulatory Commission, "Director's Status Report on Generic Activities," April 2003.
5Office of Nuclear Reactor Regulation, Nuclear Regulatory Commission, "Director's Status Report on Generic Activities," April 2003.
6 D. V. Rao, B. Letellier, C. Shaffer, S. Ashbaugh, and L. Bartlein, Los Alamos National Laboratory, "GSI-191:
Parametric Evaluations for Pressurized Water Reactor Recirculation Sump Performance," Table 5-7, "Results of Parametric Evaluations Regarding Potential for Blockage," August 2001.
7Nuclear Regulatory Commission, Generic Letter 85-22, "Potential For Loss Of Post-LOCA Recirculation Capability Due To Insulation Debris Blockage," December 3, 1985.
8Nuclear Regulatory Commission, Generic Letter 85-22, "Potential For Loss Of Post-LOCA Recirculation Capability Due To Insulation Debris Blockage," December 3, 1985.
9 Nuclear Regulatory Commission letter from John A. Grobe, Chairman - Davis-Besse Oversight Panel, to Lew W.
Myers, Chief Operating Officer, FirstEnergy Nuclear Operating Company, "Davis-Besse Nuclear Power Station NRC Integrated Inspection Report 50-346/03-17," page 39 of the enclosure, September 29,2003.
° NRC Letter dated October 7, 2003, from James L. Caldwell, Regional Administrator, to Lew Myers, Chief Operating Officer, FirstEnergy Nuclear Operating Company, "Final Significance Determination for a Yellow Finding (NRC Inspection Report 50-346/03-15) - Davis-Besse Potential Clogging of the Emergency Sump Following a Loss of Coolant Accident."
" Atomic Safety and Licensing Appeal Board, Atomic Energy Commission, Memorandum and Order (ALAB-138),
Section IV, paragraph A.l, July 31, 1973.
12 Nuclear Regulatory Commission, Information Notice 96-17, "Reactor Operation Inconsistent with the Updated Final Safety Analysis Report," March 18, 1996.
3Union of Concerned Scientists, "Safety Second: A Critical Evaluation of the NRC's First Decade," February 1985.
14 United States General Accounting Office, GAO/RCED-87-141, "Nuclear Regulation: Efforts to Ensure Nuclear Power Plant Safety Can Be Strengthened," pp. 34-35, August 1987.
15 Letter from Alex Matthiessen, Executive Director - Riverkeeper, Inc., and David Lochbaum, Nuclear Safety Engineer - Union of Concerned Scientists, to William D. Travers, Executive Director for Operations, Nuclear Regulatory Commission, "Indian Point Energy Center - Petition Pursuant to 10 CFR 2.206 - PWR Containment Sump Failure," September 8, 2003.
16 Nuclear Regulatory Commission, EA-03-086, "Issuance of Order Requiring Compliance with Revised Design Basis Threat for Operating Power Reactors," April 29, 2003.
" Nuclear Regulatory Commission from Chairman Joseph Hendrie to President Jimmy Carter, "Annual Report, 1978," page 38, February 14,1979.
18 Nuclear Regulatory Commission memo from A. W. Serkiz, Task Manager, to Distribution, "Request for Review of NUREG-0869, Revision 1," February 3, 1984.
19 Nuclear Regulatory Commission memo from A. W. Serkiz, Task Manager, to Distribution, "Request for Review of NUREG-0869, Revision 1," February 3, 1984.
20 Nuclear Regulatory Commission memo from A. W. Serkiz, Task Manager, to Distribution, "Request for Review of NUREG-0869, Revision 1," February 3, 1984.
21 Nuclear Regulatory Commission, Generic Letter 85-22, "Potential For Loss Of Post-LOCA Recirculation Capability Due To Insulation Debris Blockage," December 3, 1985.
22 Nuclear Regulatory Commission, Generic Letter 85-22, "Potential For Loss Of Post-LOCA Recirculation Capability Due To Insulation Debris Blockage," December 3, 1985.
23 Nuclear Regulatory Commission, Transcript, "Discussion of Proposed Rule on Backfitting," page 29 lines 15-25, May22, 1984.
October 29, 2003 Page 4
REGULATORY MALPRACTICE:
NRC'S "Handling" of the PWR Containment Sump Problem 24 Ivan Selin, Chairman, Nuclear Regulatory Commission, Speech before the Regulatory Information Conference, "The Future of Reactor Regulation," pp 7-8, May 9, 1995.
25 Nuclear Regulatory Commission letter from Harold R. Denton, Director - Office of Nuclear Reactor Regulation, to Chairman and Commissioners, "Compliance of NRC Licensees with NRC Regulations, Regulatory Guides, Branch Technical Positions, and Licensee Commitments," July 23, 1980.
26 Subcommittee on General Oversight and Investigations of the Committee on Interior and Insular Affairs, United States House of Representatives, "Relationship of the Nuclear Regulatory Commission to the Nuclear Industry:'
written testimony of Loren L. Bush, Jr., Chief - Program Development and Review Section, Safeguards Branch, Division of Reactor Inspection and Safeguards, Nuclear Regulatory Commission, June 11, 1987.
27 John G. Kemeny, "Saving American Democracy: The Lessons of Three Mile Island," Technology Review, June/July 1980, page 65.
28 FederalRegister, Vol. 50, No. 183, "Separate Views of Commissioner Asselstine," September 20, 1985, page 38106 29 Nuclear Energy Agency Committee on the Safety of Nuclear Installations, NEA/CNSI/R(2002)6, "Final Report:
Knowledge Base for Strainer Clogging - Modifications performed in different countries since 1992," October 3, 2002.
3 10 CFR 50.46. 139 FR 1002, Jan. 4, 1974, as amended at 53 FR 36004, Sept. 16, 1988; 57 FR 39358, Aug.31, 1992; 61 FR 39299, July 29, 1996; 62 FR 59726, Nov. 3, 1997]
October 29, 2003 Page 5
I I Brian Bennev - Supplement to petition 11 Paae 1~~~
Bennev I Brian - to Detition Suoclement Paae From: 'Dave Lochbaum" <dlochbaum @ucsusa.org>
To: <bjb~nrc.gov>
Date: 10/29/03 9:33AM
Subject:
Supplement to petition Hello Mr. Benney.
Attached are electronic copies of a letter and its enclosure that I placed in the mail to you this morning. It supplements the Riverkeeper/UCS petition.
Thanks, Dave Lochbaum Nuclear Safety Engineer Union of Concerned Scientists 1707 H Street NW Suite 600 Washington, DC 20006-3962 (202) 223-6133 x113 (202) 223-6162 fax Make your voice heard on important environmental and security issues. Join the Union of Concerned Scientists Action Network at www.ucsaction.org.
Its quick, easy, and FREE.
CC: <krabin @riverkeeper.org>
CC: