RBG-44384, Submits Response to Fuel Cladding Defect Issues Raised in 10CFR2.206 Petition.Clear Technical Basis Exists in Info Provided by River Bend Station to Deny Petition
| ML20206U793 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 02/11/1999 |
| From: | King R ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 2.206, RBF1-99-0012, RBF1-99-12, RBG-44384, NUDOCS 9902220090 | |
| Download: ML20206U793 (16) | |
Text
- _ . . . - .. . - . . ._ . - . - - .- . -_-
- e Ent:rgy operatirns,Inc.
River Bend Station 5485 U. S Highway 61 P.O. Box 220 St. Francisville, LA 70775 Te! 504 336 6225 Fax 504 635 5068 Rick J. King )
fear Safety & Regulatory Marrs February 11,1999 U.S. Nuclear Regulatory Commission Document Control Desk, OPI-37 Washington, DC 20555
Subject:
River Bend Station - Unit 1 Docket No. 50-458 License No. NPF-47 Response to Fuel Cladding Defect Issues Raised in a 10 CFR 2.206 Petition
Reference:
(1) Union of Concemed Scientists letter to L. J. Callan, " Petition Pursuant to 10 CFR 2.206, River Bend Station," dated September 25,1998.
J (2) F.J. Miraglia letter to D.A. Lochbaum," Receipt of Petition for Director's Decision Under 10 CFR 2.206," dated October 29,1998. ,
I (3) U.S. NRC letter to R.K. Edington, " River Bend Station, Unit 1 - Union of 1 Concerned Scientists Petition Submitted in Accordance with 10 CFR 2.206 Dated September 25,1998 (TAC No. MA3669)," dated December 1,1998.
(4) R.J. King letter to U.S. NRC, " Response to NRC Letter Regarding Union of Concerned Scientists' Petition Pursuant to 10 CFR 2.206," dated December 29,1998.
(5) U.S. NRC letter to R.K. Edington, dated January 26,1999.
File Nos.: G9.5 RBG-44384 4 1 "
RBF1-99-0012 f Ladies and Gentlemen:
On September 25,1998, the Union of Concerned Scientists (UCS) submitted a petition to the NRC regarding fuel cladding defects at the River Bend Station. The petition was submitted pursuant to 10 CFR 2.206. Entergy Operations, Inc. (EOI) personnel have evaluated the petition, and have determined the basis for the requested action is without technical merit. E0I requests f r N3 ~
9902220090 990211 M' PDR ADOCK 05000458T~
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.R sponse to Fuel Cladding Defect Issues Raised in a 10 CFR 2.206 Petition
- February 11,1999 '
RBF1-99-0012 RBO-44384 Page 2 of 3 that the NRC issue a Director's Decision, pursuant to 10 CFR 2.206, denying the petition in full.
The basis is summarized below, and is detailed in the attachment to this letter.
The River Bend Station continues a wrate within the bounds ofits License, including the Updated Safety Analysis Report (LoaR) and the Technical Specifications. The licensing basis is established by reading these documents as a whole, rather than by extracting selected portions.
Together, these documents contain NRC-approved limitations for operating parameters such as reactor coolant system activity, gaseous radioactive effluents, and occupational radiation exposure. These limitations provide defense-in-depth protection for the public health and safety.
The current operational conditions at River Bend Station are within the constraints of the facility licensing basis and NRC requirements, and thus the health and safety of the public remain protected.
Furthermore, the licensing basis of River Bend Station, as documented in the Technical Specifications and the USAR, adequately addresses operation with fuel cladding defects. The assumptions used in the analyses form a conservative basis for operation of River Bend, even when pre-existing fuel cladding defects are considered. An assumption used in these analyses is that the reactor coolant has levels of activity equal to Technical Specification limits, which can be attributed to pre-existing defects. Acceptance criteria for these analyses are based on the General Design Criteria (10 CFR 50, Appendix A) and the Standard Review Plan, and ensure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including anticipated operational occurrences. These acceptance criteria include provisions that such operation does not result in " loss of function of any fission product barrier l other than the fuel cladding," and that "a limited number of fuel rod cladding perforations are acceptable." Thus, analyses are not predicated upon the absence ofleaking fuel at the start of the accident. Assertions that different analyses must be performed are without technical merit.
Operation with fuel cladding defects is not inconsistent with NRC regulations. NRC regulations include limits and guidelines for protection of the health and safety of workers and the public.
Operation of River Bend Station with fuel cladding defects does not approach any of these NRC limits or guideline values. In fact, reactor coolant activity is a fraction of that permitted by the NRC. For example, licensees are required, per 10 CFR 100.11, to perfonn an accident analysis which postulates a fission product release assuming " substantial meltdown of the core with subsequent release of appreciable quantities of fission products," without exceeding the accident dose guidelines in Part 100. Obviously, the consequences of such an accident bound the l consequences of accidents involving pre-existing fuel cladding defects with significant factors of l margin. Design basis accident (DBA) analyses demonstrate that doses associated with a release l of reactor coolant during an accident (e.g., main steam line break), and having activity levels ;
significantly higher than presently being experienced at River Bend Station, would remain within '
the guideline values in 10 CFR 100. Therefore, the presence of fuel cladding defects does not pose a threat to public health and safety.
. Response to Fuel Cladding Defect Issues Raised in a 10 CFR 2.206 Petition
- February 11,1999 RBF1-99-0012 RBO-44384 Page 3 of 3 The NRC Staff correctly concluded in a letter dated October 29,1998, that the petition did not wanant immediate action by the NRC. The NRC recognized that River Bend Station took conservative actions to address the suspected condition of the fuel, and that the plant continues to operate within the limitations set forth in the Technical Specification. Attachment 1 of this letter provides a direct response to each of the petition's assertions. A clear technical basis exists in
' the information provided by River Bend Station to deny the petitien. EOI will be in attendance at the informal public hearing scheduled for February 22,1999, and would request that the NRC Staff provide River Bend Station with any new UCS information on reactor operation issued prior to the informal public hearing.
i Should the Staff have any additional questions regarding this response, please contact Ms. l Patricia Campbell at (225) 381-4615. j l
\
Sincerely, ;
)
ub .
RJK/PLC/b cc: Robert Fretz .
NRR Project Manager l U.S. Nuclear Regulatory Commission 1 M/S OWFN 13-H-3 Washington,DC 20555 NRC Resident Inspector P.O. Box 1050 St. Francisville, LA 70775 U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011
ATTACHMENT 1
. Response to Specific Assertions in the UCS Petition Dated September 25,1998 i
On September 25,1998, the Union of Concerned Scientists (UCS) submitted a petition to the NRC regarding fuel cladding defects at the River Bend Station. The petition was submitted ,
pursuant to 10 CFR 2.206. As a basis for its petition, the UCS provided quotations from the River Bend Station Updated Safety Analysis Report (USAR, or UFSAR) and other documents in ,
an attempt to demonstrate a continuing non-compliance at the River Bend Station. The NRC has asked Entergy Operations, Inc. (EOI) to provide a written response to the issues raised in this petition. The following is EOl's evaluation of each of the assenions made in the petition, with the UCS assenions followed by the River Bend Station response. Note that in each case, River Bend maintains compliance with its licensing basis, as documented in the Technicai i Specifications and the USAR. Furthermore, each UCS assenion fails to provide adequate technicaljustification as to why the petition's requested action is warranted.
- 1. PetitionAssertion 5
UFSAR 15A.2.8," General Nuciear Safety Operational Criteris," stated,"The plant shall be operated so as to avoid unacceptable consequences."
UFSAR Table 15A.2-4," Unacceptable Consequences Criteria Plant Event Category: Design Basis Accidents," defined " unacceptable consequences" as follows:
4-1 Radioactive material release exceeding the guideline values of 10CFR100. 1 4-2 Failure of the fuel barrier as a result of exceeding mechanical or thermallimits.
4-3 Nuclear system stresses exceeding that allowed for accidents by applicable industry codes, i 4-4 Containment stresses exceeding that allowed for accidents by applicable industry codes when containment is required. 1 4-5 Overexposure to radiation of plant main control room personnel.
The current operating condition at the River Bend Station apparently violates the ;
spirit, if not the letter, of Criterion 4-2 since the fuel barrier has already failed, j albeit to a limited extent. This UFSAR text does not accept a low level of fuel l barrier failure based on meeting the offsite and onsite radiation protection limits. i Integrity of the fuel barrier is an explicit criterion in addition to the radiation '
requirements. 1 River BendResponse:
l i
These two isolated citations are taken out of context; in context, they do not suppon the l conclusions drawn by the UCS, Moreover, the discussion below demonstrates that the
' Response to Fuel Cladding Defect Issues Raised in a 10 CFR 2.206 Petition February 11,1999 i RBF1-99-0012 l RBG-44834 i Attachrnent I / Page 2 of13
- two statements involve no inconsistency and that operation with the current fuel cladding failures is fully within the letter and the spirit of the River Bend licensing basis.
The first citation is correct, and is fundamental to nuclear safety. River Bend is operced "so as to avoid unacceptable consequences." In the context of the USAR (and defined
, therein - see Section 15A.2), " unacceptable consequences" are specified measures of safety - analytically determinable limits on the consequences of different classifications j of plant events - used for performing a Nuclear Safety Operational Analysis. These unacceptable consequences are defined for various plant conditions, including "Normz:
(Planned) Operation," " Anticipated (Expected) Operational Transients," " Abnormal l . (Unexpected) Operational Transients," " Design Basis (Postulated) Accidents," and
- "Special (Hypothetical) Events." Separate tables in the USAR (i.e., USAR Tables 15A.2-l 1 through 15A.2-5) identify the unacceptable consequences for each of the five plant j conditions. Unacceptable consequences are different for each of the cases.
i In their petition, the UCS focuses on a discussion of the consequences for the design basis accident. This plant condition is a highly improbable event, and safety analyses ensure that safety limits and regulatory requirements are not exceeded as a result of the
- accident occurring. This is why anicle 4-2 in USAR Table 15A.2-4 reads," Failure of a fuel barrier as a result of exceeding mechanical or thermallimits." (Emphasis added.)
- The unacceptable consequences of this type of event are independent of pre-existing fuel cladding defects. The unacceptable consequences of this event are additional fuel failures as a result of the accident occurring.
l USAR Section 15.0.3.1.1 provides further clarification in its list of unacceptable safety 4
consequences for " moderate frequency" events, which lists: " Reactor operation induced fuel-cladding failure as a direct result of the transient analysis above the minimum critical power ratio (MCPR) uncertainty level." This makes it clear that pre-existing clad defects are considered during some postulated transients. In fact, the acceptance criteria for moderate-frequency event analyses, based on the General Design Criteria (10 CFR 50, 3
Appendix A) and the Standard Review Plan, and described in the Safety Evaluation Report (SER) for River Bend Station (NUREG-0989), state the expectations for fuel cladding performance: "An incident of moderate frequency ... should not result in a loss of function of any fission product barrier other than the fuel cladding. A limited number of fuel rod cladding perforations are acceptable."
The USAR text clearly documents the acceptability of a low level of fuel cladding
- failures based on meeting the offsite and onsite radiation protection limits. For example, j USAR Table 15A.2-1 discusses the ur. acceptable consequences for normal operation.
This USAR table defines unacceptable consequences for normal operation as follows:
r, ,--r
. Response to Fuel Cladding Defect Issues Raised in a 10 CFR 2.206 Petition February 11,1999 RBF1-99-0012 P9G-44834 Attachment 1/ Page 3 of13 4-1 Release of radioactive material to the environs that exceeds the limits )
of either 10 CFR 20 or 10 CFR 50.
4-2 Fuelfailure to such an extent that were the freedfission products released to the environs via the normal discharge paths for l radioactive material, the limits of10 CFR 20 would be exceeded. l (Emphasis added.) :
4-3 Nuclear system stress in excess of that allowed for planned operation by applicable industry codes.
4-4 Existence of a plant condition not considered by plant safety analysis.
As this excerpt demonstrates, fuel cladding failures is not an unanticipated condition, and l
is an integral part of the licensing basis of River Bend. Fuel cladding defects are acceptable to the extent that they do not jeopardize radiation protection limits established in the plant Technical Specifications and other licensing basis documents. Thus, River Bend is in full compliance with the spirit and the letter of article 4-2 cited above. Given the licensing basis of the analysis methodology, USAR Table 15A.2-4 does not apply for normal operations; only USAR Table 15A.2-1 applies. Furthermore, the provisions found in USAR Table 15A.2-4 continue to be met for postulated design basis accidents.
Fmther, the USAR explicitly recognizes a level of fuel clad failures for plant normal operations. As discussed below, this is documented in USAR Chapter 11.
USAR Section 11.1 discusses that the offgas release rate of 3G4,000 Ci/sec at a 30 ,
minute delay time corresponds to design failed fuel conditions, i.e., maximum acceptable !
cladding failure for normal operation, and is conservatively based upon 105% of rated thermal power. This is consistent with Technical Specification 3.7.4, " Main Condenser Offgas," which requires that the gross gamma activity rate of the noble gases shall be
$290 mci /sec (i.e.,5290,000 pCi/sec) after decay of 30 minutes.
Note also that Technical Specification 3.4.8 "'.CS Specific Activity," establishes a coolant activity limit for I-131 of 50.2pCilg.a. As discussed in the Technical Specification Bases, the limits on the maximum allowable level of radioactivity in the reactor coolant are established to ensure, in the event of a release of any radioactive material to the environment during a DBA, radiation doses are maintained within the guidelines of 10 CFR 100. The limit for specific activity in the reactor coolant at levels that would result from significant fuel defects is an initial condition for evaluation of the consequences of a main steam line break outside containment. ,
l Thus, it is clear that operation with fuel cladding defects within acceptable activity limits is explicitly recognized and allowed under the River Bend licensing basis, including the I
1 m
Response to Fuel Cladding Defect Issues Raised in a 10 CFR 2.206 Petition February 11,1999 '
RBF1-99-0012 i RBG-44834 ;
Attachment I / Page 4 of13 !
USAR, Technical Specifications, and Technical Specification Bases. Moreover, River Bend Station operation with fuel defects does not approach these limits, and maintains adequate protection for the public health and safety. l
- 2. PetitionAssertion 1 2
UCS reviewed the UFSAR Chapter 45 description of accident analyses performed ;
for the River Bend Station. UFF at Section 15.1.1.4, " Barrier Performance," for i the loss of feedwater heating went stated: i l
The consequenco of this event do not result in any temperature or pressure l transient in excess of the criteria for which the fuel, pressure vessel, or :
containment are designed; therefore, these barrien maintain their integrity l and function as designed.
UFSAR Sections 15.1.2.4 for the feedwater controller failure - maximum event, 15.1.3.4 for the pressure regulator failure - open event, and 15.2.1.4 for the pressure regulator failure - closed event ail contain comparable statements that barrier performance was not performed because the fuel remained intact.
These analyzed events appear to be valid only when the River Bend Station is operated with no failed fuel assemblies. Operation with pre-existing fuel failures (i.e., the current plant configuration) appear to be outside of the design and licensing bases for these events.
UFSAR Section 15.4.2.5," Radiological Consequences," for the control rod withdrawal error at power event stated:
An evaluation of the radiological consequences was not made for this event since no radioactive materialis released from the fuel.
UFSAR Section 15.4.5.5," Radiological Consequences," for the recirculation flow control failure with increasing flow event stated:
As evaluation of the radiological consequences is not required for this event since no radioactive material is released from the fuel.
These analyzed events also appear valid only when the River Bend Station is operated with no failed fuel assemblies. Operation with pre-existing fuel failures (i.e., the current plant configuration) appear to be outside of the design and licensing bases for these design bases events.
Response to Fuel Cladding Defect Issues Raised in a 10 CFR 2.206 Petition February 11,1999 RBF1-99-0012 RBG-44834 Attachment I / Page 5 of 13 River BendResponse:
The information presented by UCS is taken out of context and the petition attempts to apply inappropriate and inapplicable acceptance criteria to design basis analyses. In each of the cases listed by UCS in assertion number 2, the event is classified as a " moderate frequency" event. This is also called an " anticipated operational transient," and specific criteria for unacceptable consequences are delineated in the USAR (see USAR Table 15A.2-2). For this type of anticipated transient, unacceptable performance of the fuel is described as," Reactor operation induced fuel cladding failure as a direct result of the transient analysis above the MCPR [ Minimum Critical Power Ratio] uncertainty level (0.1%)." (Emphasis added.)
The following information excerpted from the Bases for the River Bend Technical Specifications, Chapter 2.0, " Safety Limits," demonstrates how the MCPR relates to fuel cladding integrity, and how routine fuel cladding defects are already incorporated into plant operation.
Safety Limits ensure that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences. The fuel cladding integrity Safety Limit is set such that no significant fuel damage is calculated to occur if the limit is not violated. A stepback approach is used to establish the Safety Limit, such that the MCPR is not less than that specified in Technical Specification 2.1.1.2.
MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
Although some corrosion and use-related cracking may occur during the life of the cladding, cladding perforations can result from thermal stresses significantly above design conditions.
Design basis events and transients are analyzed in accordance with NRC approved methodologies, as documented in Technical Specification Section 5.6.5," Core Operating Limits Report (COLR)," and within USA.R Chapter 15. Operating limits for the MCPR are determined using this methodology to ensure that the MCPR Safety Limit is not exceeded. The purpose of the design analyses is to assess the potential for additional fuel failures during the analyzed event, and to determine acceptable operating parameters to avoid cladding perforation. The methodology provides a statistical basis for predicting that 99.9% of the fuel rods in the core avoid boiling transition, which itselfis a conservative treatment for fuel damage. Fuel damage would not nec:ssarily occur for short term boiling transition. River Bend Station is licensed to the General Electric Standard Application for Reactor Fuel (GESTAR, NEDE-24011-P-A-13), and thus the USAR inherently reflects this approved methodology. Because River Bend Station performs the analyses per the NRC approved methodology to demonstrate the
Response to Fuel Cladding Defect Issues Raised in a 10 CFR 2.206 Petition February 11,1999
. RBF1-99-0012 ,
l RBG-44834 I
'Attachinent 1/ Page 6 of13 acceptability of the MCPR analysis results, no radiological evaluation is needed.
Moreover, operation with pre-existing fuel failures is clearly within the licensing and
. regulatory design basis for the moderate frequency events identified by UCS.
- 3. PetitionAssertion i
l The effect from pre-existing fuel failures was considered, at least partially, for one !
j design bases event. UFSAR Section 15.2.4.5.1," Fission Product Release from Fuel,"
j for the main steam isolation valve closure event stated:
l J
l While no fuel rods are damaged as a consequence of this event, fission i
l product activity associated with normal coolant activity levels as well as that released from previously defective rods is released to the suppression pool as i a consequence of SRV [ safety relief valve] actuation . and vessel ;
depressurization. l The aforementioned design bases events (e.g., control rod withdrawal error at power, loss of feedwater heating, et al) are not bound by these results because the radioactive material is not " scrubbed" by the suppression pool water as it is in the i MSIV [ main steam isolation valve] closure events.
River BendResponse:
The main steam isolation valve closure event, like the events discussed in assertion number 2, above, is an anticipated operational transient. Like the other anticipated transients, unacceptable performance of the fuel is, " Reactor operation induced fuel cladding failure as a direct result of the transient analysis above the MCPR [ Minimum Critical Power Ratio] uncertainty level (0.1%)." As UCS cited,"no fuel rods are damaged as a consequence ofthis event." (Emphasis added.) This is consistent with the requirements of USAR Section 15.0.3.1.1 and of Table 15A.2-2. The presence of fuel with pre-existing clad defects is addressed by Table 15A.2-1. As discussed for assertions numbered I and 2, above, it is clear that operation with fuel cladding defects within acceptable activity limits is explicitly recognized and allowed under the River Bend licensing basis, including the USAR, Technical Specifications, and Technical Specification Bases.
With regard to the " suppression pool scrubbing" of radioactive material from the reactor coolant, the MSIV closure event is not the only moderate frequency event which results in an SRV actuation. Two of the events cited by the UCS, the Feedwater Controller Failure - Maximum Demand (USAR Section 15.1.2) and the Pressure Regulator Failure -
Closed (USAR Section 15.2.1) each result in an SRV actuation. The USAR discussion ;
i l
~ - - . - - .-.. -- -- . - - - - - - _--
' Response to Fuel Cladding Defect Issues Raised in a 10 CFR 2.206 Petition February 11,1999 RBF1-99-0012 RBG-44834
' Attachment 1/ Page 7 of13 for these events notes that radioactivity is discharged to the suppression pool, but that the activity discharged is much less than those consequences identified in USAR Section 15.2.4.5 (for the MSIV closure event). USAR Section 15.2.4.5.1, " Fission Product Release from Fuel," explains, "Since each of those transients identified previously which cause SRV actuation results in various vessel depressurization and steam blowdown rates, the transient evaluated in this section (the MSIV closure event] is that one which maximizes the radiological consequences for all transients of this nature." Thus, the USAR explicitly describes how "the aforementioned design basis events" are bound by the results for the MSIV closure event, for those events resulting in an SRV actuation.
- 4. Petition Assertion As detailed in UCS's April 1998 report on reactor operation with failed fuel cladding,it has not been demonstrated that the effects from design basis transients and accidents (i.e., hydrodynamic loads, fuel enthalpy changes, etc.) prevent pre-existing fuel failures from propagating. It is therefore possible that significantly
- more radioactive material will be released to the reactor coolant system during a j transient or accident than that experienced during steady state operation. Thus, the existing design bases accident analyses for River Bend Station do not bound its current operation with known fuel cladding failures.
River BendResponse:
In spite of quality assurance inspections by the fuel manufacturers and great care in fuel i
handling and operation by licensees, fuel cladding defects have occurred during normal operation. As described earlier, fuel cladding degradation is an anticipated condition, and is an integral part of the licensing basis of River Bend Station. Strict operating limitations for reactor coolant system activity and occupational radiation exposure limits are imposed which are sufficient to provide protection for the public health and safety.
Specified acceptable fuel design limits are applied in fuel design analyses for the explicit purpose of avoiding fuel failures during normal operation and anticipated operational occunences. Additionally, a short-term allowance for spiking activity, such as may occur as a result of an operational transient, is recognized and specified in the plant Technical Specifications. Explicit behavior of pre-existing fuel cladding defects during these events is not a required pa t of the licensing basis calculations. Instead, the event analyses are performed to determine the necessary operating limits that, when observed, will preclude fuel cladding defects from occurring, and any pre-existing clad defects are addressed and controlled through Technical Specification limits on reactor coolant system and off-gas activities. 'Ihese activity release limits conform fully to the requirements of 10 CFR 20, but also represent levels significantly greater than normally experienced during normal l
. Response to Fuel Cladding Defect Issues Raised in a 10 CFR 2.206 Petition February 11,1999 RBF1-99-0012 i RBG-44834 '
' Attachment I / Page 8 of13 steady-state operation with perforated (failed) fuel rods. Therefore, a margin exists to accommodate any additional release that may occur as a result of additional damage to existing perforated fuel rods during an anticipated operational occurrence. Additionally, l should the residual activity release consequences from the operational transient result in post-transient steady-state release rates that approach, or threaten to exceed, the plant Technical Specification limits, the plant is required to take appropriate corrective actions (such as reducing the operating power level) to assure operation remains within the plant Technical Specifications, or else the plant must shut down. In this manner, the public health and safety is assured with full conformance to the specified acceptable 10 CFR 20 requirements.
In the case of design basis transiems and accidents, analyses demonstrate the health and safety of the public remain protected if there are pre-existing fuel cladding defects, even if the defects are postulated to " grow" as a consequence of the event. For example, some design ' s accident analyses assume that all of the fission product gases in the reactor are availaole for release to the environment immediately following the event.
Two scenarios with the potential for large hydrodynamic loads and fuel enthalpy changes are the loss-of-coolant accident (LOCA) and the control rod drop accident (CRDA). The LOCA analysis results in the partial uncovering of the core and a hydrodynamic response by the plant's emergency systems. The CRDA analysis results in a localized over-power condition which can potentially cause large enthalpy changes in specific fuel rods. Each of these design basis accidents have been considered separately as they relate to the propagation of fuel cladding defects. In both scenarios, however, the propagation of a fuel failure (from the pre-existing failed fuel rod to the adjacent fuel rods) is unlikely to occur. !
> Loss of Coolant Accident The potential cladding damage mechanisms during the LOCA are (1) cladding ballooning and rupture, and (2) cladding fragmentation upon quenching after the cladding temperature excursion. Cladding ballooning occurs through outward creep deformation of the high temperature cladding caused by the condition where the fuel rod internal pressure is greater than the external pressure applied by the reactor coolant. In the failed fuel rod case, the existing cladding defect can enable equilibration between the fuel rod intemal and extemal pressure, thereby ,
eliminating the cladding ballooning driving force. Cladding fragmentation can !
occur as a result of quenching fuel which may have become brittle due to a fuel )
cladding defect (hydrided). Testing has demonstrated that no significant difference is expected between failed fuel rod and unfailed fuel rod performance during a LOCA.
l
?
I
, Response to Fuel Cladding Defect Issues Raised in a 10 CFR 2.206 Petition February 11,1999 i RBF1-99-0012 l RBG-44834 , l
'Attachinent 1/ Page 9 of13 i Local regions of the cladding in fuel rods with cladding defects can develop l essentially bulk hydride conditions prior to the onset of the LOCA. However, this ;
type oflocalized damage does not represent a departure from coolable geometry
- or provide a mechanism for failure propagation. Even in this case, the fuel rod geometry remains rod-like and coolable, with no impact on the adjacent (unfailed) l fuel rod behavior. !
j Control Rod Drop Accident
)
l The failure mechanism of the fuel cladding for this event is typically cladding ballooning and rupture. In the case of heavily hydrided cladding (due to pre-existing fuel cladding defects), pellet-cladding interaction can also occur. Similar to the conditions which exist for the LOCA analysis, depending on the extent of the pre-existing fuel cladding defect, the driving force for cladding ballooning and rupture may be precluded. Also similar to the analysis for the LOCA, the pre-existing failed fuel rod's cladding could be significantly hydrided. In this case, since the fuel rod is already failed, this condition would not lead to failure during the CRDA, but instead, the consideration is that additional cladding damage could i occur. Based on experience with failed fuel rods that have been similarly damaged due to a power increase during normal operation, and also' based on Reactivity Insertion Accident testing of fuel rods with severely hydrided cladding, failure propagation to adjacent fuel rods does not occur.
Thus, it is not necessary for the analysis methodology to explicitly account for propagation to demonstrate adequate protection of the public health and safety.
Propagation of defects from one fuel rod to an adjacent, unfailed fuel rod is unlikely to occur. j
- 5. PetitionAssertion In addition to operating with non-bounding design bases accident analyses,it appears that the River Bend licensee is also violating its licensing basis for worker .
radiation protection. UFSAR Section 12.1.1," Policy Consideration," stated:
i The purpose of the ALARA [as low as nasonably achievable] program is to maintain the radiation exposan of plant penoanel as far below the regulatory limits as is reasonably achievable. l
3 Kesponse to Fuel Cladding Defect Issues Raised in a 10 CFR 2.206 Petition j Februaiy 11,1999 RBF1-99-0012 j RBG-44834
'Attachinent 1/ Page 10 of13 ,
i UFSAR Section 12.1.2.1," General Design Considerations for ALARA Exposures,"
stated that River Bend's efforts to maintain in-plant radiation exposure as low as is reasonably achievable included:
Minimizing radiation levels in routinely occupied plant areas and in vicinity of plant equipment expected to require the attention of plant personnel.
According to the NRC Information Notice No. 87-39," Control of Hot Particle Contamination at Nuclear Plants:"
A plant operating with 0.125 percent pin-hole fuel cladding defects showed a j five-fold increase in whole-body radiation exposure rates in some areas of i the plant when compared to a sister plant with high-integrity fuel (<0.01 l percent lenkers). Around certain plant systems the degraded fuel may l elevate radiation exposure even more.
Industry experience demonstrated that reactor operation with failed fuel cladding increased radiation exposures for plant workers. The River Bend licensec has a licensing basis requirement to maintain radiation exposures for plant workers as low as reasonably achievable. The River Bend licensee informed the NRC about potential fuel cladding failures. It could shut down the facility and remove the feiled fuel assemblies from the reactor core. Instead, it continues to operate the facility with higher radiation levels.
River BendResponse:
As indicated in the US.", River Bend endorses a policy to maintain plant personnel exposure as far below regulatory limits as is reasonably achievable. River Bend is also abligated by 10 CFR 20.1101 (b), as follows: "The licensee shall use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable (ALARA)." (See also 10 CFR 50.34a and Part 50, Appendix I.) ALARA is recognized by the NRC and licensees as a guiding principle, backed up with conservative limits for exposure to workers and the public. (See 10 CFR 20.1201 and 10 CFR 20.1301.) In developing the ALARA principle, the NRC recognized that "as low as reasonably achievable"is not a radiation protection standard.
(40 Fed. Reg. 19439(1975)).
River Bend Station ALARA Program l River Bend implements an ALARA program using a multi-tiered program consisting of policies and procedures. For example, a Site Policy entitled "ALARA" establishes
Response to Fuel Cladding Defect Issues Raised in a 10 CFR 2.206 Petition !
February 11,1999 RBF1-99-0012 RBG-44834
' Attachment I / Page 11 of13 Entergy's commitment to comply with the intent of applicable codes, standards, and othe.
criteria accepted by the industry for assuring the safe and efficient operation of River
)
Bend Station. The operating philosophy for maintaining the occupational and collective radiation exposure to station personnel and to the general population ALARA is funher ,
defined in this pohey. 4 l
As required by the plant policy, procedures are developed to implement the ALARA program at RBS. Some of these procedures cover topics such as temporary shielding control, posting of radiologically controlled areas, radiological surveys, radiation work permits, personnel monitoring, control and handling of radioactive material, access control, and a radiation protection training and qualification program. There is also an overall administrative procedure for the ALARA Program, as well as an upper-tier document for River Bend Station discussing the Radiation Protection Plan. These procedures, in concert with operations, maintenance group, and work planning procedures and processes, work to implement a radiation protection program at RBS that evaluates, plans, and provides instruction in the conduct of maintenance and routine operations such that dose received individually and collectively is maintained ALARA. 1 As stated in the Rimr Bend USAR (section 12.1.2.1 General Design Considerations for ALARA Exposu es), "The general design considerations and methods employed to maintain in-plant radiation exposure ALARA have two objectives: Minimizing the amount of time plant personnel spend in radiation areas; and Minimizing radiation levels in routinely occupied plant areas and in the vicinity of plant equipment expected to ;
require the attention of plant personnel."
This objective is satisfied since routinely occupied areas of the plant have not seen a sustained increase in general area dose rates. When entry is required into some specific areas of the off-gas system, River Bend Station implements controls such as continuous monitoring of dose rates by a radiation protection technician. The radiation protection department has been and continues to monitor radiological conditions very closely in the j areas which may be potentially impacted.
l As a course of normal business, each entry into an area with a potential for personnel to receive dose requires that they log onto a radiation work permit with a pre-determined limit for dose received and dose rate encountered on the entry. In addition, the individual doses received and overall station collective dose is monitored on a daily basis to track actual dose received against that which has been estimated for operations and scheduled maintenance activities.
, Response to Fuel Cladding Defect Issues Raised in a 10 CFR 2.206 Petition
' February 11,1999 i I
RBF1-99-0012 I RBG-44834
' Attacliment 1/ Page 12 of13 Industry Experience >
NRC Information Notice (IN) No. 87-39, " Control of Hot Particle Contamination at i Nuclear Plants," was pwyared after a case study into the operation and refueling practices ;
of a pressurized water reactor (PWR) (see NUREG/CR-4485). River Bend Station is not a i PWR, and thus correlation ofimpact on systems and radiological characteristics with those at River Bend Station is impractical. Even the cases referring to fuel element defects cited in IN 87-39, for the most part, are plants which operated for a substantial l period of time with significant fuel element failures. The current situation at River Bend does not approach a fuel failure of the magnitude described in the Information Notice regarding these industry events. Thus, the information contained in IN 87-39 is oflittle '
relevance to River Bend Station at this time.
ALARA and Operational Regulatory Constraints NRC guidance for meeting the ALARA principle is clear that the bounding limits !
authorized in plant Technical Specifications are consistent with meeting ALARA. See 10 ;
CFR Part 50, Appendix I. Operation with fuel cladding defects at current levels are well within River Bend Station Technical Specification limits. The NRC's regulation (10 CFR Part 50, Appendix I, Sec. IV) states that a licensee is:
... permitted the flexibility of operations, compatible with considerations i of public health and safety, to assure that the public is provided a ,
dependable source of power even under urusual coraditions which may '
temporarily result in releases higher than numerical guides for design i objectives but still within levels that assure that the average population exposure is equivalent to small fractions of doses from natural background radiation." l Thus, the NRC recognizes that, under conditions far more severe than those being experienced at River Bend Station, licensees may continue to operate and still comply )
with ALARA criteria and objectives (the "ALARA Principle"). l l
The suggestion that River Bend shutdown solely to address fuel cladding defects is l contrary to the ALARA principle. River Bend maintains a low radiation dose to its plant j workers during operation as the result of an effective ALARA program and a well- ,
operating plant. River Bend is currently averaging approximately 0.12 person-rem per l day. If River Bend Station were to shutdown solely to remove fuel bundles which exhibit !
_ fuel cladding defects which are not exhibiting individual degradation,6e exposure to
. workers would be greatly in excess of this low exposure value. This outage womd be in
- t i
, pesponse to Fuel Cladding Defect Issues Raised in a 10 CFR 2.206 Petition February 11,1999 RBF1-99-0012 '
RBG-44834
" Attacliment 1/ Page 13 of13 addition to the normal refueling outage scheduled for April 1999. Continued operation within the bounds of Technical Specifications is consistent with the ALARA principle.
- 6. PetitionAssertion 1
- Since it appears that operation with one or more failed fuel assemblies is not permitted by its design and licensing bases, River Bend must immediately shut down. The facility must remain shutdown until
O The River Bend licensee removes the failed fuel assemblies from the reactor
-OR- '
, D The River Bend licensee properly updates the plant's design and licensing bases to permit the plant to operate with known fuel damage.
River BendResponse:
As detailed in the responses above, the design and licensing bases of River Bend do permit operation with fuel clad defects, provided that the plant continues to meet certain regulatory and administrative requirements. Furthermore, River Bend Station has not exceeded any Technical Specification limits for reactor coolant system (RCS) specific activity (see Technical Specification 3.4.8, "RCS Specific Activity") or for Main ,
Condenser Offgas activity releases (see Technical Specification 3.7.4). River Bend i Station has taken reasonable measures to identify the extent of fuel cladding defects and to take appropriate corrective actions. River Bend Station continues to conservatively operate within the bounds ofits Operating License and USAR, and any requirement to modify such documents is unwarranted.
CONCLUSION The NRC has previausly denied the UCS any enforcement action to require the immediate shutdown of River Bend Station in order to address potential fuel cladding defec'.s. Having now demonstrated that the individual assertions of the UCS petition are without technical merit, do not involve safety significance or impact public health and safety, and do not provide any basis for modifying, su. spending or revoking the Piver Bend Operating License, we respectfully request that the NRC issue a final Director's Decision, pursuant to 10 CFR 2.206, to deny the petitioner's requests in full.
t