ML20236T114
| ML20236T114 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 07/23/1998 |
| From: | Roche M GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC |
| Shared Package | |
| ML20135G042 | List: |
| References | |
| NUDOCS 9807270440 | |
| Download: ML20236T114 (13) | |
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GPU Nuclear, Inc.
Oyster Creek Nuclear Generating Station Facility License No. DPR-16 m
Technical Specification Change Request No. 260 Docket No. 50-219 Applicant hereby submits a proposed change to Appendix A Technical Specification page 2.1-1.
By:
Michael B. Roche Vice President and Director Oyster Creek Sworn and subscribed to before me this 23rd day of July 1998.
A Notary Public ofNJ 9807270440 980723 PDR ADOCK 06000219 P PDR
United States of America .
., + Nuclear Regulatory Conunission .
.In the Matter of )
) Docket No. 50-219 GPU Nuclear, Inc. )
Certificate of Service
- This is to certify that a copy of Technical Specification Change Request No. 260 for the Oyster Creek Nuclear Generating Station Operating License, filed with the U.S. Nuclear Regulatory Commission on July 23,1998 has this day of July 23,1998, been served on the Mayor of Lacey Township, Ocean County, New Jersey by deposit in the United States mail, addressed as follows:
The Honorable Louis A. Amato, Jr.
Mayor ofLacey Township 818 West Lacey Road Forked River, NJ 08731 l
By:
Michael B. Roche Vice President and Director Oyster Creek 3
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1 Attachment]
Oyster Creek Technical Specification Change Request No. 260 GE Affidavit Concerning Proprietary Information in Attachment 1
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i GE Nuclear Energy GenwalEbectrt Cornpany l P o. Bat 780. Wiltrangton. NC 28402 l
1 Affidavit I, Glen A. Watford, being dgy sworn, depose and state as follows:
(1) I am Manager, Nuclear Fuel Engineering, General Electric Company ("GE") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to ,
be withheld, and have been authorized to apply for its withholding. j (2) The information sought to be withheld is contained in Attachment I to letter number 1940 ,
20391, Technical Specification Change Request No. 260. The proprietary text has identifying brackets in the right hand margin.
1 (3) In making this application for withholding of proprietary information of whicn it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom ofInformation Act ("FOIA"),
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5 USC Sec. 552(b)(4), and the Trade Secrets Act,18 USC Sec.1905, and NRC regulations 10 j CFR 9.17(a)(4) and 2.790(a)(4) for " trade secrets tnd commercial or financial information !
obtained from a person and privileged or con &dential" (Exemption 4). The material for which !
exemption from disclosure is here sought is all " con 5dential commercial information," and some portions also qualify under the narrower dennition of" trade secret," within the meanings assigned i to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Enerny Pro _ieet v.
Nuclear Regulatory Commission. 975F2d871 (DC Cir.1992), and Public Citizen Health Research Groun v. FDA. 704F2dl280 (DC Cir.1983).
(4) Some examples of categories of information which fit into the definition of proprietary information are:
- a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other l companies; )
- b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the derign, manufacture, shipment, installation, assurance of quality, or licensing of a similar product; f
- c. Information which reveals cost or price information, production capacities, budget le,els, or commercial strategies of General Electric, its customers, or its suppliers; .
- d. Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential commercial value to General Electric; ,
- c. Information which discloses patentable subject matter for which it may be desirable to l obtain patent protection, j l
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The information sought to be withheld is considered to be proprietary for the reasons set forth in both paragraphs (4)a. and (4)b., above.
(5) The information sought to be withheld is being submitted to NRC in con &dence. The inrormation is of a sort customarily held in confidence by GE, and is in fact so held. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in (6) and (7) following. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GE, no public disclosure has been made, and it is not av:u)ble in public sources. All disclosures to third patties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.
(6) Initial approval of proprietcy treatment of a document is made by the manager of the originating component, the person moet likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.
(7) he procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the infonnation, and then only in accordance with appropriate regulatory provisions or proprietary agreements.
(8) The information identi5ed in paragraph (2) is classi5ed as proprietary because it contains details of GE's Safety Limit MCPR ana:ysis and the corresponding mJts which GE has applied to Oyster Creek's actual core design with GE's fuel.
The development of the methods used in these analyses, along with the testing, development and approval of the supporting methodology was achieved at a significant cost, on the order of several million dollars, to GE.
(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities.
The stability analysis is part of GE's con prehcasive BWR safety and technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.
The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by GE.
The precise value of the expertise to devise an evaluation process and apply the correct analytical ;
methodology is difficult to quantify, but it clearly is substantial.
GE's competitive advantage will be lost ifits competitors are able to use the results of the GE cxperience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.
em.c m m.a.
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'Ibe value of this information to GE would be lost if the information were disclosed to the public.
Making such information available to competitors without their having been required to undertake
- a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive L
GE of the opportunity to exercise its competitive advantage to seek an adequate return on its large l- investment in developing these very valuable ana!>tical tools.
State ofNorth Carolina )
County of New Hanover ) SSg Glen A. Watford, being duly sworn, deposes and says:
That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.
Executed at Wilmington, North Carolina, day of. this h d .19N Glen A.W rd General Electric Company l
Subscribed and nyom before me this N day of Mala- . 19 6
(/ 0
, kW Notary Public, State of North Carolina My Commission Expires /e/,r/,248 /
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cNucwunw. manum Page 3
Attachment 3 Oyster Creek Nuclear Generating Station Technical Specification Change Request No. 260 Non-Proprietary Version L: Channe Reauested ,
The note pertaining to Technical Specification 2.1.A will be revised to indicate the applicability of the specification to operating Cycle 17. The proposed change is contained
. on page 2.1-1. A mark-up of the existing page is in Attachment 4. The revised page is in l
Attachment 5.
IL Discussion of Proposed Channes The Oyster Creek fuel cladding integrity safety limit or safety limit minimum critical power ratio (SLMCPR) was originally a' generic value for each fuel design calculated by the fuel L
vendor, General Electric (GE). It was determined that the generic analysis had become non-conservative due to changes in fuel design and core loading that invalidated certain assumptions regarding fuel bundle local power peaking and core radial peaking factors used in the generic analysis As a result, GE revised the SLMCPR analysis methodology and until it is approved generically a plant-specific calculation to determine the SLMCPR value must be performed for each operating cycle. Oyster Creek Technical Specification 2.1.A regarding SLMCPR was approved for operating Cycle 16 via License Amendment No.192 dated August 26,1997. This license amendment request will apply the fuel
- cladding safety limit to Cycle 17.
IIL Safety Assessment The fuel cladding integrity safety limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting -
in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate
. boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit.
Operating limits are specified to maintain adequate margin to the onset of the boiling transition. The parameter used for core design and monitoring is the critical power ratio.
The critical power ratio (CPR) is defined as the ratio of the critical power (bundle power
! at which some point in the fuel assembly experiences onset of boiling transition) to the operating bundle power.
The critical power is determined at the same mass flux, inlet temp::rature and pressure that exist at the specified reactor condition. Thennal margin is stated in terms of the minimum
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Attachment 3 Page 2 of 6 value of the critical power ratio (MCPR), which corresponds to the most limiting fuel assembly in the core. To ensure that adequite margin is maintained, a design requirement based on a statistical analysis was selected. This requirement states that more than 99.9 percent of the fuel rods would be expected to avoid boiling transition during moderate frequency transients, allowing for uncertainties in manufacturing and monitoring the core operating state. Both the safety limit and normal operating limit for the fuelin terms of MCPR are derived fromjhis basis.
The Cycle 17 SLMCPR was calculated to be 1.08. The current Technical Specification limit of 1.09 is conservative to the 1.08 value and the Technical Specification value of 1.09 will remain unchanged. General Electric calculated the SLMCPR based on input supplied by GPU Nuclear. The Technical Specification limit for a minimum critical power ratio of 1.49 ta provide margin to the MCPR safety limit in the event of reactor thermal-hydraulic instability is not impacted The adequacy of the 1.08 value for Cycle 17 is based on the following:
Control Rod Pattern Development fbr the Ovster Creek Cycle 17 SLMCPR Analysis Projected control rod patterns for the rodded burn through the Cycle were used to deplete the core to the cycle exposures to be analyzed. At the desired cycle exposures, the bundle exposure distributions and their associated R-factors, determined in accordance with Reference 1, were utilized for the SLMCPR cases analyzed. The use of different rod patterns to achieve the desired cycle exposure has been shown to have a negligible impact on the actual calculated SLMCPR. An estimated SLMCPR was obtained for an exposure point at the beginning of cycle (BOC), peak hot excess (PHE) and end of cycle (EOC) in order to establish which exposure point (s) would produce the highest (most conservative) calculated SLMCPR.
1 For each cycle exposure point ofinterest, the SLMCPR is analyzed with radial power distributions that maximize the number of bundles at or near the operating limit CPR during rated power operation. This approach satisfies the stipulation in Reference 1 that l the number of rods susceptible to boiling transition be maximized. GE has established criteria to determine if the control rod pattern and resulting radial power distribution are acceptable. These criteria were discussed with the NRC inspection team during a May 6-10,1996 inspection and han since been incorporated into the GE technical design procedures. These criteria include no gross violations of Technical Specification operating limits (e.g., MCPR, Maximum Average Planar Linear Heat Generaticr Rite, Linear Heat Generation Rate), criticality (calculated, normalized kar near 1.0) and total number of bundles within [GE PROPRIETARY INFORMATION REMOVED] of the MCPR core.
Subsequently, GE ins replaced the last criteria with the MCPR importance parameter (MIP). The MIP value is a measure of the flatness of the MCPR distribution. The value of MIP decreases as the core MCPR distribution becomes more peaked in order to place f
l bundles near the operating limit MCPR. A lower bound on the value of MIP is specified
Attachment 3 Page 3 of 6 to assure that the MCPR distribution used will provide sufficient conservatism in the value of SLMCPR calculated from that distribution.
Different rod patterns were analyzed until the criteria on the above parameters were met.
Starting from a defined set of patterns, known from prior experience, to yield the flattest possible MCPR distributions narrowed the rod pattern search. This was done for all three-cycle exposures, bje ynning-of-cycle (BOC), peak hot excess point (PHE) and end-of-cycle (EOC). The EOu exposure is actually performed between EOC - 1000 MWD /ST and EOC - 2000 MWD /ST in order to meet the criteria specified above and be able to place sufficient fuel bundles near the MCPR operating limit. A Monte Carlo analysis was then performed for each of the three exposure points to establish the !
maximum SLMCPR for the cycle.
Comparison of Ovster Creek Cycle 17 SLMCPR versus the Generic GE9B Valae Table I summarizes the relevant input parameters and results of the SLMCPR detennination for both the generic GE8X8NB and the Oyster Creek Cycle 17 core.
GESTAR II (Reference 2) specifies that the SLMCPR analysis for a new fuel design shall be performed for a large high power density plant assuming a bounding equilibrium core.
The GE9B product line generic SLMCPR (1.06) was determined according to this specifiestion.
The Oyster Creek Cycle 17 core is near equilibrium with 532 GE9B out of 560 fuel bundles. The 28 bundles that are not GE9B are GE8B (GE8X8EB) located on the core periphery and the limiting MCPR locations are all GE9B. The SLMCPR for the generic analysis occurs at peak hot excess reactivity (4.4 GWD/MT). The Oyster Creek core SLMCPR occurs at EOC-1500 GWD/MT. This differen :e is attributable to core design and fuel bundle design differences between Oyster Creek end the generic analysis. The lower peak bundle power for the Oyster Creek analysis is due to the higher operating CPR ,
limit (1.53) for Oyster Creek than used in the generiu analysis (1.30). This does not (
impact the calculation of the SLMCPR.
In general, the calculated safety limit is dominated by two key parameters: (1) flatness of the core bundle-by-bundle MCPR distributions and (2) flatness of the bundle pin-by-pin )
power /R-factor distributions. The Oyster Creek Cycle 17 analysis produces both a flatter !
bundle-by-bundle and pin-by-pin power distributions. In Table 1, the number of bundles I thht are within [GE PROPRIETARY INFORMATION REMOVED]ACPR is greater for !
the Oyner Creek Cycle 17 analysis than the generic analysis. The more bundles closer to limits yields more rods susceptible to boiling transition and a higher SLMCPR. The MIP i
value for EOC 17 is 2.59, well above the required minimum MIP value of 1.0 for cycle specific evaluations, further supporting the degree of flatness for the MCPR distribution.
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Attachment 3 Page 4 of 6 The Oyster Creek analysis has a pin-by-pin distribution that is flatter than the generic analysis because the generic analysis includes a number of controlled bundles that are within [GE PROPRIETARY INFORMATION REMOVED) of MCPR while the Oyster Creek analysis does not include any controlled bundles within [GE PROPRIETARY INFORMATION REMOVED). The R-factor distribution for controlled fuel bundles is more peaked than the uncontrolled R-factor distribution. By having all uncontrolled fuel bundles within [GE PROPRIETARY INFORMATION REMOVED] of the limiting MCPR, the Oyster Creek analysis will have a flatter pin-by-pin distribution. This will also yield more rods susceptible to boiling transition and a higher SLMCPR.
Therefore, the Oyster Creek Cycle 17 analysis provides a more conservative (flatter) power distribution than the generic analysis and the higher SLMCPR for Oyster Creek Cycle 17 analysis than the generic analysis (1.08 vs.l.06)is reasonable. The current Oyster Creek SLMCPR of 1.09 bounds the 1.08 value for Cycle 17 and provides a conservativelimit.
IV. Information Suncorting a Findine of No Significant Hazards Consideration GPU Nuclear has concluded that the proposed change, verifying adequacy of the current SLMCPR for operating Cycle 17, does not involve a Significant Hazards Consideration. i In support of this determination, an evaluation of each of the three standards set forth in 10 CFR 50.92 is provided below.
- 1. The croposed TS chance does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The derivation of the Cycle 17 SLMCPR for Oyster Creek for incorporation into the TS, and its use to determine cycle-specific thermal limits, has been performed using NRC-approved methods. Additionally, interim implementing procedures, .
which incorporate cycle-specific parameters, have been used. Based on the use of these calculations, the Cycle 17 SLMCPR of 1.09 will not increase the probability or consequences of an accident.
The basis of the MCPR Safety Limit calculation is to ensure that greater than 99.9% of all fuel rods in the ccre avoid transition boiling if the limit is not violated.
A SLMCPR of 1.09 preserves adequate margin to transition boiling and fuel damage in the event of a postulated accident. The probability of fuel damage is not increased.
Therefore, the proposed TS change does not involve an increase in the probability or consequences of an sccident previously evaluated.
- 2. The proppsed TS change does not create the p.pssibility of a new or dif1'erent kind of accident from any accident previousiv evalugigd.
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Attachment 3 Page5of6 The MCPR Safety Limit is a Technical Specification aumerical value designed to t ensure that fuel damage from transition boiling does not occur as a result of the limiting postulated accident. The limit cannot create the possibility of any new type of accident. The Cycle 17 SLMCPR has been calculated using NRC-approved methods. Additionally, interim procedures, which incorporate _ cycle-specific parameters, have been used. Therefore, the proposed TS change does not create the possib35ty'of a new or different kind of accident, from any accident previously evaluated.
- 3. The orooosed TS change does not involve a significant reduction in a margin of safety.
The margin of safety as defined in the TS Bases will remain the same. The Cycle 17 SLMCPR is calculated using NRC-approved methods, which are in accordance with the current fuel design and licensing criteria. Additionally, interim implementing procedures, which incorporate cycle-specific parameters, have beei i used. The MCPR Safety Limit remains high enough to ensure that greater than 99.9% of all fuel rods in the core will avoid transition boiling if the limit is not l '
violated, thereby preserving fuel cladding integrity. The: fore, the proposed TS change does not involve a reduction in a margin of safety.
3 V. Information Sunnortine an Environmental Asmsment An environmental assessment is not required for the proposed change since the proposed change conforms to the criteria for " actions eligible for categorical exclusion" as specified in 10 CFR S t.22(c)(9). The proposed change will have no impact on the environment.
The proposed change does not involve a significant ha'ards consideration as discussed in the preceding section. The proposed change does not involve a significant change in the types or significant increase in the amounts of any effluents that may be released off-site.
In addition, the proposed change does not involve a significant increase in individual or cumulative occupational radiation exposure.
VL Conclusion
- The proposed change to the TS, which confirms the existing SLMCPR for Cycle 17, has been reviewed in accordance with Section 6.5 of the Oyster Creek Technical ~
j Specifications and it has been concluded there are no unreviewed safety questions. As i discussed above, using the standards in 10 CFR 50.92, GPU Nuclear believes that there are no Significant Hazards Considerations involved with the proposed change.
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Attachment 3 Page 6 of 6 Table 1 Comparison of Generic GE9B and Oyster Creek Cycle 17 Cores Quantity, description GE9B Oyster Creek Generic Cycle 17 Number ofBundles in Core 764 560 _
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References:
- 1) Licensing Topical Report, General Electric BWR Thennal Analysis Basis (GETAB): Data, Correlation and Design Application, NEDO-10958-A, January 1977
- 2) General Electric Nuclear Energy Document NEDE-24011-P-A-11. General Electric Standard Application for Reactor Fuel (GESTAR II), dated November 1995 L
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l Attachinent 4 l Oyster CreekTechnical Specification Change Request No. 260 l
Mark-up Revision to Technical Specifications l
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