ML023600466
| ML023600466 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 12/16/2002 |
| From: | Gallagher M AmerGen Energy Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2130-02-20330 | |
| Download: ML023600466 (25) | |
Text
SAmerGensm AmerGen Energy Company, LLC wwwexeloncorp corn An Exelon/British Energy Company 200 Exelon Way Suite 345 Kennett Square, PA 19348 10 CFR 50.90 December 16, 2002 2130-02-20330 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Oyster Creek Generating Station Facility Operating License No. DPR-16 NRC Docket No. 50-219
Subject:
Technical Specification Change Request No. 312 - Safety Limit Minimum Critical Power Ratio
References:
- 1. AmerGen Letter dated June 26, 2002; Re: "Technical Specification Change Request No. 291 - Safety Limit Minimum Critical Power Ratio"
- 2. AmerGen Letter dated August 1, 2002; "Response to Request for Additional Information -Technical Specification Change Request No. 291, Safety Limit Minimum Critical Power Ratio (TAC NO. MB5505)"
- 3. NRC Letter dated September 26, 2002; "Oyster Creek Nuclear Generating Station Issuance of Amendment Re: Cycle 19 Safety Limit Minimum Critical Power Ratio (TAC NO. MB5505)"
In accordance with 10 CFR 50.4(b)(1), enclosed is Technical Specification Change Request No. 312.
The purpose of this Technical Specification Change Request is to revise Oyster Creek Technical Specifications to incorporate revised Safety Limit Minimum Critical Power Ratio (SLMCPR) values due to a revision of the cycle specific analysis performed by Global Nuclear Fuel (GNF) for Oyster Creek Cycle 19. The original GNF analysis for Cycle 19 was included in Technical Specification Change Request No. 291, which was approved in Technical Specification Amendment 233 dated September 26, 2002. The revised SLMCPR values, based on the enclosed GNF revised analysis for Oyster Creek Cycle 19, are 1.10 (three loop operation) and 1.09 (for both four or five loop operation).
The evaluation supporting this Technical Specification Change Request is contained in Enclosure 1 to this letter, and the proposed marked up Technical Specification pages are contained in Enclosure 2. (letter from T. G. Orr (Global Nuclear Fuel) to K. Donovan (Exelon Generation Company, LLC), dated September 13, 2002) specifies the revised SLMCPR values for Oyster Creek Cycle 19. contains information proprietary to Global Nuclear Fuel. Accordingly, it is requested that be withheld from public disclosure. Enclosure 4 provides a non-proprietary version of the Global Nuclear Fuel document. An affidavit certifying the basis for this application for withholding as required by IOCFR2.790(b)(1) is provided as Enclosure 5.
Using the standards in 10 CFR 50.92, AmerGen Energy Company, LLC (AmerGen) has concluded that these proposed changes do not constitute a significant hazards consideration, as described in the enclosed evaluation performed in accordance with 10 CFR 50.9 1(a)(1). Pursuant to 10 CFR 50.91(b)(1), a copy of this Technical Specification Change Request is provided to the designated official of the State of New Jersey, Bureau of Nuclear Engineering, as well as the Chief Executive of the township in which the facility is located.
This proposed change to the Technical Specifications has undergone a review in accordance with Section 6.5 of the Oyster Creek Technical Specifications. No new regulatory commitments are established by this submittal.
NRC approval of this change is requested by May 23, 2003. This requested approval date is to allow the revised SLMCPR values to be implemented in time to provide thermal margin relief that is needed beginning at mid cycle (June 2003).
If any additional information is needed, please contact David Robillard at (609) 971-4793.
I declare under penalty of perjury that the foregoing is true and correct.
Sincerely, 12/ /n02 Ex ec u-te d O n
Enclosures:
Michael P. Gallagher Director, Licensing & Regulatory Affairs Mid-Atlantic Regional Operating Group (1)
Oyster Creek Technical Specification Change Request No. 312 Evaluation of Proposed Changes (2)
Oyster Creek Technical Specification Change Request No. 312 Markup of Proposed Technical Specification Page Changes (3)
Letter from T. G. Orr (Global Nuclear Fuel) to K. Donovan (Exelon Generation Company, LLC), dated September 13, 2002, Proprietary Version (4)
Letter from T. G. Orr (Global Nuclear Fuel) to K. Donovan (Exelon Generation Company, LLC), dated September 13, 2002, Non-Proprietary Version (5)
Global Nuclear Fuel Affidavit Certifying Request for Withholding from Public Disclosure cc:
H. J. Miller, Administrator, USNRC Region I P. S. Tam, USNRC Senior Project Manager, Oyster Creek R. J. Summers, USNRC Senior Resident Inspector, Oyster Creek File No. 02079
cc:
Mr. Kent Tosch, Director Bureau of Nuclear Engineering Department of Environmental Protection CN 415 Trenton, NJ 08628 The Honorable Louis Amato Mayor of Lacey Township 818 West Lacey Road Forked River, NJ 08731
ENCLOSURE 1 Oyster Creek Technical Specification Change Request No. 312 Evaluation of Proposed Changes 2130-02-20330 Page 1 of 6
1.0 INTRODUCTION
This letter is a request to amend Operating License No. DPR-16.
The proposed changes would revise the Operating License to incorporate the revised Safety Limit Minimum Critical Power Ratio (SLMCPR) for three loop operation and four or five loop operation due to the cycle specific analysis, as revised, performed by Global Nuclear Fuel for Oyster Creek Cycle 19. This change supports Cycle 19 operation. NRC approval of this change is requested by May 23, 2003 in order to allow the revised SLMCPR values to be implemented in time to provide thermal margin relief that is needed beginning at mid cycle (June 2003).
AmerGen Energy Company, LLC (AmerGen) requests the following changed replacement pages be inserted into the existing Technical Specifications:
Revised Technical Specification Pages: 2.1-1, 2.1-2, and 2.1-3 The marked up pages showing the requested changes are provided in Enclosure 2.
2.0 DESCRIPTION
OF PROPOSED AMENDMENT The proposed amendment involves revising the Safety Limit Minimum Critical Power Ratio (SLMCPR) values contained in Technical Specification 2.l.A (page 2.1-1) from 1.12 to 1.10 for three recirculation loop operation and from 1.11 to 1.09 for both four and five recirculation loop operation. The current SLMCPR values were determined for Oyster Creek based on the reload core design for Cycle 19. As approved in Technical Specification Amendment 233, the current SLMCPR values were determined in accordance with NRC approved methodology described in "General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A (GESTAR II),
Amendment 25.
Amendment 25 provides the methodology for determining the cycle specific MCPR safety limits. Amendment 25 was also used in determining the revised Cycle 19 SLMCPR values, and it is intended to use Amendment 25 for determining future SLMCPR values. The NRC safety evaluation approving Amendment 25 is contained in a letter from the NRC to General Electric dated March 11, 1999, [F. Akstulewicz (NRC) to G. A. Watford (GE), "Acceptance for Referencing of Licensing Topical Reports NEDC-32601P, Methodology and Uncertainties for Safety Limit MCPR Evaluations; NEDC-32694P, Power Distribution Uncertainties for Safety Limit MCPR Evaluations; and Amendment 25 to NEDE-2401 1-P-A on Cycle Specific Safety Limit MCPR," (TAC Nos. M97490, M99069 and M97491)].
This proposed amendment is to change the SLMCPR values for Cycle 19 based on the use of an approved revised methodology documented in NEDC-32694P-A "Power Distribution Uncertainties for Safety Limit MCPR Evaluations", whereas the previous evaluation used the older General Electric BWR Thermal Analysis Basis (GETAB) methodology for treating power distribution uncertainty. Technical Specification pages 2.1-2 and 2.1-3 are revised to add the approved revised methodology for SLMCPR analysis.
2130-02-20330 Page 2 of 6
3.0 BACKGROUND
The current Cycle 19 SLMCPR values were determined in accordance with the NRC approved methodology described in "General Electric Standard Application for Reactor Fuel," NEDE-2401 1 P-A-14 (GESTAR-II), and U. S. Supplement, NEDE-2401 1-P-A-14-US, June, 2000, which incorporates Amendment 25. Amendment 25 provides the methodology for determining the cycle specific MCPR safety limits. Amendment 25 was also used for determining the revised Cycle 19 SLMCPR values. The NRC safety evaluation approving Amendment 25 is contained in a letter from the NRC to General Electric Company, dated March 11, 1999 [F. Akstulewicz (NRC) to G.
A. Watford (GE), "Acceptance for Referencing of Licensing Topical Reports NEDC-32601P, Methodology and Uncertainties for Safety Limit MCPR Evaluations; NEDC-32694P, Power Distribution Uncertainties for Safety Limit MCPR Evatuation; and Amendment 25 to NEDE 24011-P-A on Cycle Specific Safety Limit MCPR," (TAC Nos. M97490, M99069 and M97491)].
The current Cycle 19 SLMCPR values were approved in Technical Specification Amendment 233 dated September 26, 2002.
Subsequent to the approval of the current Cycle 19 SLMCPR values, it was decided to pursue a revised analysis using the approved revised methodology in NEDC-32694P-A. The approved revised methodology yields lower SLMCPR values primarily due to an improved treatment of the power distribution uncertainty that reduces the conservatism of the GETAB method of power allocation. The Cycle 19 core loading and the uncertainty values used in the revised SLMCPR analysis are the same as in the original GETAB based analysis submitted in Technical Specification Change Request 291.
4.0 REGULATORY REQUIREMENTS & GUIDANCE 10 CFR 50, Appendix A, General Design Criteria (GDC) 10 requires that the reactor core and associated coolant, control, and protective systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences. Safety limits are required to be included in the Technical Specifications by 10 CFR 50.36. The SLMCPR is developed to assure compliance with GDC 10 for fuel cladding integrity. The SLMCPR ensures sufficient conservatism in the operating MCPR limit that, in the event of an anticipated operational occurrence from the limiting condition of operation, at least 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties. The SLMCPR is recalculated each refueling cycle due to fuel replacement.
5.0 TECHNICAL ANALYSIS
The proposed Technical Specification change will revise Technical Specification 2.1.A to reflect the cycle specific analysis, as revised, performed by Global Nuclear Fuel for Oyster Creek Cycle 19.
The current Cycle 19 SLMCPR values were determined in accordance with the NRC approved methodology described in "General Electric Standard Application for Reactor Fuel," NEDE-2401 1-2130-02-20330 Page 3 of 6 P-A-14 (GESTAR-II), and U. S. Supplement, NEDE-2401 1-P-A-14-US, June, 2000, which incorporates Amendment 25. Amendment 25 provides the methodology for determining the cycle specific MCPR safety limits. Amendment 25 was also used for determining the revised Cycle 19 SLMCPR values. The NRC safety evaluation approving Amendment 25 is contained in a letter from the NRC to General Electric Company, dated March 11, 1999.
GNF has revised the SLMCPR analysis for Cycle 19 by applying the improved methodology documented in NEDC-32694P-A whereas the previous analysis used the older GETAB methodology for treating the power distribution uncertainty. The revised analysis results in lower SLMCPR values primarily due to an improved treatment of the power distribution uncertainty that reduces the conservatism of the GETAB method of power allocation.
The revised analysis establishes SLMCPR values that will ensure that at least 99.9% of all fuel rods in the core avoid transition boiling if the limit is not exceeded. The SLMCPR values are calculated to include cycle specific parameters, which include 1) the actual core loading, 2) conservative variations of projected control blade patterns, 3) the actual bundle parameters (e.g.
local peaking), and 4) the full cycle exposure range. The Cycle 19 core loading and the uncertainty values used in the revised SLMCPR analysis are the same as in the original GETAB based analysis approved in Technical Specification Amendment 233. The revised SLMCPR values for Cycle 19 are 1.10 (three loop operation) and 1.09 (for both four loop and five loop operation) as shown in. Additional information regarding the cycle specific SLMCPR values for Oyster Creek Cycle 19 is contained in the previously approved TS Amendment 233 (TAC No. MB5505) and the revised GNF analysis contained in Enclosure 3.
The analyses performed demonstrate the proposed change is acceptable since no fuel thermal limits or other licensing basis acceptance criteria are adversely affected.
Conclusion The proposed changes to implement revised SLMCPR values for Oyster Creek Cycle 19 provide safety limit protection by ensuring that at least 99.9% of the fuel rods in the core will not experience boiling transition, which complies with the requirements of 10 CFR 50 Appendix A, GDC-10 regarding acceptable fuel limits. The proposed safety limit values have been developed by Global Nuclear Fuel using plant and cycle specific fuel and core parameters in accordance with NRC approved methodologies. Consequently, the proposed Technical Specification changes will not adversely affect nuclear safety or safe plant operations.
6.0 REGULATORY ANALYSIS
10 CFR 50.36(c)(1) requires that safety limits be included in the plant Technical Specifications.
Therefore, the SLMCPR is included in the Oyster Creek Technical Specifications. The SLMCPR values have been determined in accordance with NRC approved methodology described in "General Electric Standard Application for Reactor Fuel," NEDE-2401 1-P-A-14 (GESTAR-II),
and U. S. Supplement, NEDE-2401 1-P-A-14-US, June 2000.
2130-02-20330 Page 4 of 6 In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
7.0 NO SIGNIFICANT HAZARDS CONSIDERATION AmerGen has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The derivation of the cycle specific Safety Limit Minimum Critical Power Ratio (SLMCPR) values for incorporation into the Technical Specifications (TS), and their use to determine cycle specific thermal limits, has been performed using the methodology discussed in "General Electric Standard Application for Reactor Fuel, "NEDE-2401 1-P-A-14 (GESTAR-II), and U.S. Supplement, NEDE 2401 1-P-A-14-US, June, 2000, which incorporates Amendment 25. Amendment 25 was approved by the NRC in a safety evaluation report dated March 11, 1999.
The basis of the SLMCPR calculation is to ensure that at least 99.9% of all fuel rods in the core avoid transition boiling if the limit is not violated. The revised SLMCPR values developed in the revised analysis preserve the existing margin to transition boiling and fuel damage in the event of a postulated accident. The proposed safety limit values have been developed by Global Nuclear Fuel using plant and cycle specific fuel and core parameters in accordance with NRC approved methodologies. Neither the probability nor the consequences of fuel damage will be increased as a result of this change.
The proposed changes to the Technical Specification Bases are considered administrative only and have no affect on nuclear safety or safe plant operations.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The SLMCPR is a Technical Specification numerical value, designed to ensure that transition boiling does not occur in greater than 99.9% of all fuel rods in the core if the limit is not violated.
The revised SLMCPR values are calculated using NRC approved methodologies discussed in 2130-02-20330 Page 5 of 6 "General Electric Standard Application for Reactor Fuel," NEDE-2401 1-P-A-14 (GESTAR-II), and U.S. Supplement, NEDE-2401 1-P-A-14-US, June, 2000, which incorporates Amendment 25.
The SLMCPR is not an accident initiator, and its revision will not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes to the Technical Specification Bases are considered administrative only and have no affect on nuclear safety or safe plant operations.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
There is no significant reduction in the margin of safety previously approved by the NRC as a result of the proposed change to the SLMCPR values. The revised SLMCPR values are calculated using methodology discussed in "General Electric Standard Application for Reactor Fuel," NEDE 2401 1-P-A-14 (GESTAR-II), and U.S. Supplement, NEDE-2401 1-P-A-14-US, June, 2000, which incorporates Amendment 25. The SLMCPR values ensure that at least 99.9% of all fuel rods in the core will avoid transition boiling if the limit is not violated when all uncertainties are considered, thereby preserving the fuel cladding integrity. The margin of safety, as defined in the Technical Specifications, for all events is maintained.
The proposed changes to the Technical Specification Bases are considered administrative only and have no affect on nuclear safety or safe plant operations.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, AmerGen concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
8.0 ENVIRONMENTAL CONSIDERATION
A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
2130-02-20330 Page 6 of 6 Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
9.0 PRECEDENT The proposed SLMCPR changes for Oyster Creek are similar to the SLMCPR changes recently approved for Quad Cities I in Amendment No.210, dated November 14, 2002. The revised Oyster Creek Cycle 19 SLMCPR analysis was performed by Global Nuclear Fuel using plant and cycle specific fuel and core parameters, and NRC approved methodologies including NEDO-1095 8-A, General Electric BWR Thermal Analysis Basis (GETAB), NEDC-32694P, Power Distribution Uncertainties for Safety Limit MCPR Evaluation, and Amendment 25 to NEDE-240 11-P-A (GESTAR H).
10.0 REFERENCES
a) NEDE-2401 1-P-A-14 "General Electric Standard Application for Reload Fuel," (GESTAR II), and U.S. Supplement, NEDE-2401 1-P-A-14-US, June 2000 b) NRC Safety Evaluation Report dated March 11, 1999 (F. Akstulewicz (NRC) to G. A.
Watford (GE), "Acceptance for Referencing of Licensing Topical Reports NEDC-32601P, Methodology and Uncertainties for Safety Limit MCPR Evaluations; NEDC-32694P, Power Distribution Uncertainties for Safety Limit MCPR Evaluation; and Amendment 25 to NEDE 24011-P-A on Cycle Specific Safety Limit MCPR," (TAC Nos. M97490, M99069, and M97491).
c) NEDE-24011-P-A-14, (GESTAR II) Amendment 25.
d) Letter from T. G. Orr (Global Nuclear Fuel) to K. Donovan (Exelon Generation Company, LLC) dated September 13, 2002 (Proprietary).
ENCLOSURE 2 Oyster Creek Technical Specification Change Request No. 312 Markup of Proposed Technical Specification Page Changes Revised TS Pages 2.1-1 2.1-2 2.1-3
SECTION 2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMIT - FUEL CLADDING INTEGRITY Applicability:
Applies to the interrelated variables associated with fuel thermal behavior.
Obiective:
To establish limits on the important thermal hydraulic variables to assure the integrity of the fuel cladding.
Specifications:
A.
When the reactor pressure is greater thann r equal to 800 psia and the core flow is greater than or equal to 10% of rated e existence of a minimum CRITICAL POWER RATIO (MCPR) less than 1lfor both four or five loop operation and j./0 ;,4-Z for three loop operation shall constitute violation of the fuel cladding integrity safety limit.
B.
When the reactor pressure is less than 800 psia or the core flow is less than 10% of rated, the core thermal power shall not exceed 25% of rated thermal power.
C.
In the event that reactor parameters exceed the limiting safety system settings in Specification 2.3 and a reactor scram is not initiated by the associated protective instrumentation, the reactor shall be brought to, and remain in, the COLD SHUTDOWN CONDITION until an analysis is performed to determine whether the safety limit established in Specification 2.1.A and 2.1..B was exceeded.
D.
During all modes of reactor operation with irradiated fuel in the reactor vessel, the water level shall not be less than 4'8" above the TOP OF ACTIVE FUEL.
Amendment No.: 75,135,192, 202, 218,228, 233)
OYSTER CREEK 2.1-1
Bases:
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. However, the uncertainties in monitoring the core operating state and in the procedure used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity safety limit is defined as the CRITICAL POWER RATIO in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.
The Safety Limit MCPR(1) is determined using the General Electric Thermal Analysis Basis, GETAB(2), which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power' The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality*
(X) - Boiling Length (L), GEXL, correlation.
I The use of the GEXL correlation is not valid for the critical power calculations at pressures below 800 psia or core flows less than 10% of rated. Therefore, the fuel cladding integrity safety limit is protected by limiting the core thermal power.
At pressures below 800 psia, the core elevation pressure drop (0 power, 0 flow) is greater than 4.56 psi. At low power and all flows this pressure differential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and all flows will always be greater than 4.56 psi. Analyses show that with a flow of 28 x 103 Ibs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, bundle flow with a 4.56 psi driving head will be greater than 28 x 103 lbs/hr irrespective of total core flow and independent of bundle power for the range of bundle powers of concern. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors this corresponds to a core thermal power of more than 50%. Thus, a core thermal power limit of 25% for reactor pressures below 800 psi or core flow less than 10% is conservative.
Plant safety analyses have shown that the scrams caused by exceeding any safety setting will assure that the Safety Limit of Specification 2.1.A or 2.1.B will not be exceeded. Scram times are checked periodically to assure the insertion times are adequate. The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g.,
scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause fuel damage. Specification 2.1.C requires that appropriate analysis be performed to verify that backup protective instrumentation has prevented exceeding the fuel cladding integrity safety limit prior to resumption of POWER OPERATION. The concept of not approaching a, Safety Limit provided scram signals are OPERABLE is supported by the extensive plant safety analysis.
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k cer4eU-c OYSTER CREEK 2.1-2 Amendment No.: 75, 192, 228, 233 Pedoces fLw c-01serYvi17YsK e&42e G&TAY3 &,egod o-P pouj-r a 116 c a7oA. Al
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If reactor water level should drop below the TOP OF ACTIVE FUEL, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. With a water level above'the TOP OF ACTIVE FUEL, adequate cooling is maintained and the decay heat can easily be accommodated. It should be noted that during power generation there is no clearly defined water level inside the shroud and what actually exists is a mixture level. This mixture begins within the active fuel region and extends up through the moisture separators. For the purpose of this specification water level is defined to include mixture level during power operations.
The lowest point at which the water level can presently be monitored is 4'8" above the TOP OF ACTIVE FUEL. Although the lowest reactor water level limit which ensures adequate core cooling is the TOP OF ACTIVE FUEL, the safety limit has been conservatively established at 4'8" above the TOP OF ACTIVE FUEL.
REFERENCES (1)
NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel (GESTAR II)
(latest approved version as specified in the COLR)
(2)
General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design I
Application, NEDO-10958-A, January 1977.
C4)
NF-Dc C-3WPO I ?- 'A Amendment No.: 75,135, 192, 228, 233)
OYSTER CREEK 2.1-3 7ý
ENCLOSURE 4 Letter from T.G. Orr (Global Nuclear Fuel) to K. Donovan (Exelon Generation Company, LLC), dated September 13,2002 Non-Proprietary Version
Attachment Additional Information Regarding the September 13,2002 Cycle Specific SLMCPR for Oyster Creek Cycle 19 References
[1]
Letter, Frank Akstulewicz (NRC) to Glen A. Watford (GE), "Acceptance for Referencing of Licensing Topical Reports NEDC-32601 P, Methodology and Uncertainties for Safety Limit MCPR Evaluations; NEDC-32694P, Power Distribution Uncertainties for Safety Limit MCPR Evaluation; and Amendment 25 to NEDE-24011 P-A on Cycle Specific Safety Limit MCPR," (TAC Nos. M97490, M99069 and M97491), March 11, 1999.
[2]
Letter, Thomas H. Essig (NRC) to Glen A. Watford (GE), "Acceptance for Referencing of Licensing Topical Report NEDC-32505P, Revision 1, R-Factor Calculation Method for GEJ1, GEl2 and GEl3 Fuel," (TAC No. M99070 and M95081), January 11, 1999.
[3]
General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, NEDO-10958-A, January 1977.
[4]
Letter, GA. Watford (GNF) to J.E. Donoghue (NRC), Final Presentation Material for GEXL Presentation - February 11,2002; FLN-2002-004; February 12, 2002.
[5]
GNF-A design record file (DRF) 0000-0004-7201 titled "Oyster Creek Cycle 19 Safety Limit MCPR (SLMCPR)".
[6]
Letter, Michael P. Gallagher (Exelon) to U.S. NRC Document Control Desk,
'Technical Specification Change Request No. 291 - Safety Limit Minimum Critical Power Ratio for Oyster Creek Generating Station, June 26, 2002, 2130-02-20153.
Comparison of Oyster Creek SLMCPR Values for Cycle 19 Table 1 summarizes the relevant input parameters and results of the SLMCPR determination for the Oyster Creek C cle 19 core. The current SLMCPR evaluation and the evaluation performed previously["' were both performed using NRC approved methods and uncertainties".6]. The current evaluation uses the revised methodology documented in NEDC-32694P-A1 l' whereas the previous evaluation used the older GETAB131 methodology.
Both evaluations used the Oyster Creek specific uncertainties indicated in Table 2. The current evaluation yields lower calculated SLMCPR values than those calculated previously[6) primarily because the revised methodology eliminates the artificial correlating of the four bundles surrounding a TIP string that occurs with the GETAB methodology. (See Section 4.3 ofNEDC-32601P-A for more detail.)
These calculations are specific to Oyster Creek, Cycle 19. The cycle-specific quantities that have been shown to have some impact on the determination of the safety limit MCPR (SLMCPR) are provided in Table 1. Comparisons to Cycle 18 SLMCPR values have been provided previously in Reference 6.
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
(( ]I Page 1 of 6
((I]1
Attachment Additional Information Regarding the September 13, 2002 Cycle Specific SLMCPR for Oyster Creek Cycle 19 power/R-factor distributions. Greater flatness in either parameter yields more rods susceptible to boiling transition and thus a higher calculated SLMCPR.
(( ))
The uncontrolled bundle pin-by-pin power distributions are also used in the Oyster Creek Cycle 19 calculations. Pin-by-pin power distributions are characterized in terms of R-factors using the NRC approved methodology1 21.
For the Oyster Creek Cycle 19 limiting case analyzed at EOC, ((1]
Summary The calculated 1.09 Monte Carlo SLMCPR for Oyster Creek Cycle 19 is consistent with what one would expect for the revised methodology ((
)) the 1.09 SLMCPR value is appropriate.
Based on all of the facts, observations and arguments presented above, it is concluded that the calculated SLMCPR value of 1.09 for the Oyster Creek Cycle 19 core is appropriate. It is reasonable that this value is 0.02 lower than the 1.11 value calculated previously. This value applies to both four and five loop operation.
For three loop operations (3LO) the calculated safety limit MCPR for the limiting case is 1.10 as determined by specific calculations for Oyster Creek Cycle 19.
Supporting Information The following information is provided in response to NRC questions on similar submittals regarding changes in Technical Specification values of SLMCPR. Only those items that require a plant/cycle specific response are presented below since all the others are contained in the references that have already been provided to the NRC.
For Oyster Creek Cycle 19 the specific power distribution uncertainties that are applied are even higher than the GETAB1 31 values as indicated in Table 2. Use of higher specific values is provided for in the NRC SERI'].
The core loading information for Oyster Creek Cycle 19 is provided in Figure 1. The core loading for Cycle 18 was provided previously in Reference [6]. The impact of the fuel loading pattern differences on the calculated SLMCPR is correlated to the values of (( ))
(( ))
Page 2 of 6
(( ))
Attachment Additional Information Regarding the September 13, 2002 Cycle Specific SLMCPR for Oyster Creek Cycle 19 Table 1 Comparison of the Oyster Creek Cycle 19 SLMCPR Data QUANTITY, DESCRIPTION Oyster Creek Cycle 19 Number of Bundles in Core 560 Limiting Cycle Exposure Point EOC Cycle Exposure at Limiting Point [MWd/STU]
10400 Reload Fuel Type GEI I Latest Reload Batch Fraction [%]
37.2 %
Latest Reload Average Batch Weight %
3.70 %
Enrichment Batch Fraction for GE9 66.1%
Batch Fraction for GEl 1 33.9%
Core Average Weight % Enrichment 3.54 %
Core MCPR (for limiting rod pattern) 1.55 Power distribution uncertainty See Table 2, Column 2 Non-power distribution uncertainty See Table 2, Column 2 Calculated Safety Limit MCPR 1.09
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Page 3 of 6
((1]1
Attachment Additional Information Regarding the September 13,2002 Cycle Specific SLMCPR for Oyster Creek Cycle 19 Table 2 Comparison of Oyster Creek Cycle 19 Specific Inputs to NRC-accepted Values COLUMN I COLUMN 2 DESCRIPTION Uncertainty Values (%)
Oyster Creek previously accepted by Specific Values NRC
(%)
Non-power Distribution From Table 2.1 of Uncertainties NEDC-32601P-A Core flow rate (derived from pressure 2.5 TLO 2.5 TLO drop) 6.0 SLO 6.0 SLO Individual channel flow area
(( ))
(())
Individual channel friction factor 5.0 5.0 Friction factor multiplier
(( ))
((I]
Reactor pressure
((1]
((1]
Core inlet temperature 0.2 0.2 Feedwater temperature
(( ))
(( ]
Feedwater flow rate
(( ))
((I]
Power Distribution Uncertainties GETAB uncertainties as Specific Values consistent with the Revised used to produce values
(%)
Methodology of NEDC-32601P-A shown in Table 4.1 of NEDC-32601P-A GEXL R-factor
((1))
((]
Random effective TIP reading 1.2 TLO 1.2 TLO 2.85 SLO 2.85 SLO Systematic effective TIP reading
((1]
(())
Integrated effective TIP reading
(())
((1]
Bundle power
(( ]I
(( ))
Effective total bundle power
(( ]I
(())
uncertainty
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Page 4 of 6
(( ))
Attachment Additional Information Regarding the September 13,2002 Cycle Specific SLMCPR for Oyster Creek Cycle 19 Figure 1 Reference Core Loading Pattern - Cycle 19 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 18 Al fIDIC GFE G IF-Il CI, E Gf E L
1CIEI G
D U
I B
AE E
E E
E C
C F
A 8
rA A
DIFIC F
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El A IE E.
A E
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A A1 9
AC CEBECE A
E B
E B
B E
B E
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10 ADCFECEAE A
E C
E E
C E
A EAECE FCA 11 A C C F A E C E B E
C E
A A
E C
E B
E C E A F C C A
12ACFECEA E
C E
C E
E C
E C
EAECE FCDA 13 AD CF BE A E B E A
E B
B E
A E
BE A
E B F CD A
14 A
D C F B E A E B E
A E
B 8
E A
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15 A
D C F E C E A
E C
E C
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A E C E F C D A 16 A
C F A E C E B E
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17A D E
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18 A C C E B E C E A E
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E B
A E C E B E C C A
19 EAAIAEC A
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ict IUAFI4P A
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i -
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17 19 21 23 25 27 29 31 33 35 19 20 21 22 23 24 25 26 1
3 5
7 9
11 13 15 Number in Core 52 50 48 46 42 40 38 36 34 32 30 28 26 24 22 20 18 16 14 12 10 8
6 4
2 43 45 47 49 51 Cycle Loaded A
GE9B-P8DWB348-12GZ-80U-145-T6 140 17 B
GE9B-P8DWB338-1 1GZ-80U-145-T6 40 17 C
GE9B-P8DWB348-12GZ-80U-145-T6 136 18 D
GE9B-P8DWB338-11GZ-80U-145-T6 48 18 E
GE1 1 -P9HUB369-12GZ-1 OOT-1 45-T6-2560 144 19 F
GE1 I -P9HUB374-13GZ-1 OT-1 45-T6-2559 46 19 G
GE9B-P8DWB348-12GZ-80U-145-T6 6
19 Total 560
(( ))
Page 5 of 6 R[ ))
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D A IA IA-
iAttachment Prepared by:
Additional Information Regarding the September 13,2002 Cycle Specific SLMCPR for Oyster Creek Cycle 19 Verified by:
"HL/VSf-J.P. Rea Technical Program Manager I. Maldonado Technical Program Manager
]1
]1
((
Page 6 of 6
ENCLOSURE 5 Global Nuclear Fuel Affidavit Certifying Request for Withholding from Public Disclosure
Global Nuclear Fuel A Joint Venture of GE. Toshiba, & Hitachi Affidavit I, Glen A. Watford, state as follows:
(1) I am Manager, Fuel Engineering Services, Global Nuclear Fuel - Americas, L.L.C.
("GNF-A") 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.
(2) The information sought to be withheld is contained in the attachment, "Additional Information Regarding the Cycle Specific SLMCPR for Oyster Creek Cycle 19,"
September 13, 2002.
(3) In making this application for withholding of proprietary information of wlhich it is the owner or licensee, GNF-A relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4) and 2.790(a)(4) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all "confidential commercial information," and some portions also qualify under the narrower definition of "trade secret," within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA 704F2d1280 (DC Cir. 1983).
(4) Some examples of categories of information which fit into the definition of proprietary informaition are:
- a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GNF-A's competitors without license from GNF-A constitutes a competitive economic advantage over other companies;
- b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;
- c. Information which reveals cost or price information, productiotn capacities, budget levels, or commercial strategies of GNF-A, its customers, or its suppliers;
- d. Information which reveals aspects of past, present, or future GNF-A customer funded development plans and programs, of potential commercial vhlue to GNF A;
- e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.
Page 1
Affidavit The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b., above.
(5) The information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GNF-A, 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 GNF-A, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties 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 proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GNF-A. Access to such documents within GNF-A is limited on a "need to know" basis.
(7) The 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 GNF-A are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.
(8) The information identified in paragraph (2) is classified as proprietary because it contains details of GNF-A's fuel design and licensing methodology.
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 GNF-A or its licensor.
(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GNF-A's competitive position and foreclose or reduce the availability of profit making opportunities. The fuel design and licensing methodology is part of GNF-A's comprehensive 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 GNF-A or its licensor.
I d'EUicensngkaflidavit\\lgnfaffa davu doc Page 2
Affidavit 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.
GNF-A's competitive advantage will be lost if its competitors are able to use the results of the GNF-A experience 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.
The value of this information to GNF-A 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 GNF-A of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.
I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.
Executed at Wilmington, North Carolina, this 13th day of September
,2002.
Glen A. Watford Global Nuclear Fuel - Americas, LLC I *\\NEihcensing'aflidavitVgnfa_aff'davit doe Page 3