ML20211M245
| ML20211M245 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 07/16/1997 |
| From: | Storz L Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20046D868 | List: |
| References | |
| LCR-H97-05, LCR-H97-5, LR-N97433, NUDOCS 9710140103 | |
| Download: ML20211M245 (15) | |
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JUL 161997 LR-N97433 LCR H97-05 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555' I
Gentlemen:
REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS (SUPPLEMENT)
SAFETY LIMIT MINIMUM CRITICAL POWER RATIO (SIJ4CPR)
HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 5C-354 On March 31, 1997, Public Service Electric & Gas (PSE&G) Company transmitted, via letter LR-N97187, a proposed change to the Hope Creek Technical Gpecifications (TS). The proposed changes revised TS 2.1.2, " THERMAL POWER, High Pressure and High Flow",
ACTION a.l.c for LCO 3.4.1.1, " Recirculation Loops" and the Bases for TS 2.1, " Safety Limits". The changes contained in that request implemented an appropriately conservative Safety Limit Minimum Critical Power Ratio (SLMCPR) for Hope Creek's Cycle 7 (current cycle) core and fuel designs. Justification for those proposed changes was developed from General Electric SLMCPR analyses performed to address SLMCPR issues identified in a 10CFR21 notification made by General Electric on May 24, 1996.
This letter supplements the original TS change request by providing revised SLMCPR values for the upcoming operating cycle (Cycle 8). The justification for these proposed changes is being revised to include Cycle 8 specific analyses and the 10CFR50.92 evaluation is being appropriately updated to reflect the new changes. The proposed changes have been evaluated in accordance with 10CFR50. 91 (a) (1) , using the criteria in 10CFR50.92 (c), and a determination has been made that this request involves no significant hazards considerations.
The basis for the requested change is provided in Attachment 1 to this letter. A 10CFR50.92 evaluation, with a determination of no significant hazards consideration, is provided in Attachment 1 The marked up Technical Specification pages affected by the proposed changes are provided in Attachment 3. Pursuant to Attachment 4 of this letter, this submittal contains proprietary THE ATTACHMENTS TO Ti" *ETTER CONTAIN PROPRIETARY INFORMATION
- NO ,R PUBLIC DISCLOSURE -
(h 9710140103 971003 PDR ADOCK 05000354 P PDR i
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Document Control oesk -2 JUL 16 897 LR-N97433 information and therefore should be withheld from public disclosure. Attachment 5 of this letter provides additional information relative to the Hope Creek Cycle 7 SL!iCPR analyses.
Upon NRC approval of this proposed change, PSE&G requests that the amendment be made effective on the date of issuance, but allow an implementation period of sixty days to provide sufficient time for associated' administrative activities. A copy of this submittal has,been transmitted to the State of New Jersey.
Should you have any questions regarding this request, we W 11 be pleased to discuss them with you.
Sincerely,
/ )M'O
- Affidavit Attachments (5)
THE ATTACHMENTW TO THIS LETTER CONTAIN PROPRIETARY INEVRMATION
- NOT PVR PUBLIC DISCLOSURE -
a
9 Documer.t Control Desk jg{ g g gg7 LR-N97433 C Mr. H. Miller, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allandale Road King of Prussia, PA 19406 Mr. D. Jaffe, Licensing Project Manager - HC l
U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike l'
Mai) Stop 14E21 Rockville, MD 20852 Mr. S. Morris (X24)
USNRC Senior Resident Inspector - HC Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering 33 Arctic Parkway CN 415 Trenton, NJ 08625 t
THE ATTACHMENTS TO THIS LETTER CONTAIN PROPRIETARY INFORMATION
- NOT FOR PUBLIC DISCLOSURE -
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Document Control Desk' dOb10 NOI LR-N97433 JPP BC Senior Vice President _ Nuclear Engineering- (N19)
General Manager - Hope Creek Operations (H07)
Director - QA/NSR (X01)
Manager - Business Planning & Co-Owner Affairs (N18)
Manager - Hope Creek Operations (H01)
Manager - System Engineering - Hope Creek (H18)
Manager.- Nuclear Review Board (N38)
Manager - Licensing & Regulation (N21)-
Manager - Hope Creek Licensing (N21)
Supervisor - Hope Creek Fuels (N20)
J. J. Keenan, Esq. (N21)
Records Management (N21)
Microfilm Lopy Files Nos. 1.2.1 (Hope Creek), 2.3 (LCR H97-05)
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I
REF: LR-N97433 LCR H97-05 STATE OF NEW JERSEY )
) SS.
COUNTY OF SALEM )
L. F. Stor:, being duly sworn according to law deposes and says:
I am Senior Vice President - Nuclear Operations of Public Servicu
- Electric and Gas Company, and as such, I find the matters set forth in the above referenced letter, concerning Hope Creek l
Generating Station, Unit 1, are true to the best of my knowledge, information and belief, h '
(44t < 6 o
Subscribed and Sworn to before me this /6 day of hlu , 19"7 Y
r)4 Lut. 0 s Notary Public of New Jersey BAtlARA A.POELL N0TAgt PW ON M My Commission expires on uvC mM WCE e v eesta THE ATTACHMENTS TO THIS LETTER CONTAIN PROPRIETARY INFORMATION
- NOT FOR PUBLIC DISCLOSURE -
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, ~. Document Control D0ck LR-N97433 Attcchment 1 LCR H97-05 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET No. 50-354 SAFETY LIMIT MINIMUM CRITICAL POWER RATIO (SIJ4CPR) CHANGES BASIS FOR REQUESTED CHANGE:
- The changes proposed in this re' quest implement an appropriately ,
conservative SLMCPR for the Hope Creek Cycle 8 core and fuel designs. These changes are required to address SLMCPR issues i
identified in a 10CFR21 notification made by General Electric on May 24, 1996 (Reference 1). That 10CFR21 notification discussed non-conservative SLMCPR calculation methodologies that impacted Hope Creek. As a result of the issues discussed in that 10CFR21 notification, Hope Creek issued Licensee Event Report (LER) 96-014-00, dated May 14, 1996 (which was supplemented by LER 96-014-l 01, dated September 30, 1996). As described in the corrective actions in that LER, Hope Creek is conservatively controlling the minimum critical power ratio (MCPR) at an interim value that bounds initial accident conditions. NRC approval of the changes l proposed in this submittal, will provide resolution of this issue
' and support required Technical Specification (TS) changes needed j for Cycle 8 operation.
REQUESTED CHANG 5' AND PURPOSE:
As shown in Attachment 3 of this letter, TS 2.1.2 is being modified to: 1) replace the 1.07 MCPR limit for two recirculation loop operation with a 1.10 MCPR limit (for Cycle
- 8); and 2) replace the 1.08 MCPR limit for single recirculation 4 loop operation with a 1.12 MCPR limit (for Cycle 8). In i addition, the Bases for TS 2.1, " Safety Limits", will be revised to reflect the new 1.10 MCPR limit for two recirculation loop i operation and 1.12 MCPR limit for single recirculation loop operation. These changes will also require a revision to ACTION a.1.c fcr LCO 3.4.1.1 to reference the proposed 1.12 MCPR limit for single loop operation.
i THE ATTACHMENTS TO THIS LETTER CONTAIN PROPRIETARY INFORMATION
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Page 1 of 8
. Oscument Control DOsk- LR-N97433 Attcchment 1 LCR H97 1 BACKGROUNL: ,
In the course of calculating a cycle-specific SLMCPR for another utility, General Electric Company (General Electric) determined !
that the GESTAR II (General Electric Standard Application for !
2 Reactor Fuel, NEDE-240ll-P-A-11 , and U. S.-Supplement NEDE-
- 2 4 0 l l- P- A- 11 -US;, November 17, 1995) generic SLMCPR may be non-conservative when applied to some core and fuel designs. _The NRC was informed of this condition in a telephone call by General Electric on March 27, 1996, which became the subject of a 10 CFR Part 21 notification from General Electric dated May 24, 1996 (Reference 1).
When this issue was identified to Hope Creek, LER S6-014-00 was transmitted to the NRC to document this issue. Subsequently, the Hope Creek SLMCPR values were confirmed for the current operating Cycle 7 in a letter from General Electric to PSE&G, dated May 8, 1996. As an interim compensatory measure, Hope Creek has implemented administrative controls to ensure that the operating ,
MCPR limit bounds initial accident analysis conditions.
Since that time, General Electric has calculated a revised plant-specific SLMCPR value for Hope Creek Cycle 7 as part of the : '
Reload Licensing Analysis. This calculated SLMCPR value for Hope Creek Cycle 7 was based upon NRC approved methods (General Electric Standard Applica tion eor Reactor Fuel, NEDE-24011-P-A-11, and U. S. Supplement NEDE-24011-P-A-ll-US, November 17, 1995) which have been discussed between General Electric and the NRC during meetings _ held on April 17, 1996 and May 6 through 10, 1996. The implementing procedures are identical to those used for similar recent analyses for other facilities and described in General Electric's proposed Amendment 25 to GESTAR II (R. J. Reda (GE) to T. E. Collins (NRC), Proposed Amendment 25 to GE
. Licensing Topical Report NEDE-24011-P-A (GESTAR II) on Cycle Specific Safety Limit MCPR, December 13, 1996). These procedures incorporate cycle specific parameters into the analysis which include: 1) the actual core loading; 2) conservative variations of projected control blade patterns; 3) the actual-bundle
' Revision iI has since been superseded by Revision 13. dated August,19% The Revision 13 material peninent to this application is unchanged from Revision 11. For purposes related to evaluation of this application. Revisions 11 and 13 may be considered equivalent and used interchangeably.
THE ATTACHMENTS TO THIS LETTER CONTAIN PROPRIETARY INFORMATION
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Document Control Dosk LR-N97433 Attcchment 1 LCR H97-05 parameters; and 4) the full cycle exposure range. This calculation resulted in Cycle 7 SLMCPR values of 1.08 for two loop operation and 1.09 for single loop operation. On March 31, 1997, Public Service Electric & Gas (PSE&G) Company transmitted, via letter LR-N97187, License Change Request H97-05 to revise the Hope Creek TS to incorporate these Cycle 7 results.
Subsequently, General Electric has performed analysis for the i
Hope Creek Cycle 8 core and fue'l design. The method used to analyze Cycle 8 and determine the new SLMCPR values is provided in the following section. PSE&G proposes that the Hope Creek Technical Specifications be revised as indicated in Attachment 3 of this submittal to incorporate these new SLMCPR values for Cycle 8 operation.
JUSTIFICATION OF REQUESTED CHANGES:
The proposed changes contained in this submittal will revise the Technical Specifications to reflect the new SLMCPR values calculated by General Electric for Hope Creek. As stated previously, these plant specific evaluations were performed by General Electric for Hope Creek, Reload 7, Cycle 8 and were calculated using NRC approved methods.
Introduction For Hope Creek, the Fuel Cladding Integrity Safety Limit is set such that no mechanistic 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 the onset of transition boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that the onset of transition 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 procedures 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 CPR in the limiting fuel assembly for which mor2 than 99.9% of the fuel rods in the core are expected THE ATTACHMENTS TO THIS LETTER CONTAIN PROPRIETARY INFORMATION
- NOT FOR PUBLIC DISCLOSURE -
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. Document Control D sk LR-N97433 Attochment 1 LCR H97-05 to avoid boili.ng transition considering the power distribution within the core and all uncertainties. The new SLMCPRs for Cycle 8 at Hope Creek are 1.10 for two loop operation and 1.12 for single loop operation.
Control Rod Pattern Development for the Hope Creek Cycle 8 SLMCPR Analysis Pro]ected control blade 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 i distributions and their associated R-factors were utilized fer
! the SLMCPR cases to be analyzed. The use of different rod patterns to achieve the desired cycle exposure has been shown to have a negliglole impact on the actual calculatod SLMCPR. An estimated SLMCPR was obtained for an exposure point near beginning of cycle (BOC), middle of cycle (MOC), and the end of cycle (EOC) in order to establish which exposure points v.ould produce the highest (most conservative) calculated SLMCPR.
The Safety Limit MCPR is analyzed with radial power distributions that maximits the number of bundles at or near the Operating i Limit MCPR during rated power operation. This approach satisfies the stipulation in Reference 2 that the number of rods susceptible to boiling transition be maximized. General Electric has established criteria to determine if the control rod patterns and resulting radial power distributions are acceptable. These criteria were discussed with the NRC inspection team during the May 6-10, 1996 inspection and have since been incorporated into the General Electric technical design procedures. These criteria include no gross violations of technical specification operating limirs (e.g., MCPR, MAPLHGR, LHGR), criticality (calculated, normalized k.te near one) and total number of bundles M of the MCPR of the core.
Different rod patterns were analyzed until the criteria on the above parameters were met. The rod pattern search was narrowed by starting from a defined set of patterns known from prior experience to yield the flattest possible MCPR distributions.
This was done for three exposure points in the cycle. A Monte Carlo analysis was then performed for the to establish THE ATTACHMENTS TO THIS LETTER CONTAIN PROPRIETARY INFORMATION NOT FOR PUBLIC DISCLOSURE -
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. Document Control DOsk LR-N97433 Attachment 1 LCR H97-05
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the maximum SLMCPR for the cycle. The maximum SLMCPR occurred at Comparison of Hope Creek Cycle 8 SLMCPR versus the Generic GE9B Value Table 1 summarizes the relevant input parameters and results of the SLMCPR determination for both the generic GE9B core and the
- Hope Creek Cycle 8 core. GESTAR II (Reference 3) 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 C-lattice GE9B product line generic SLMCPR (1.07) was determined according to this specification. Hope Creek Cycle 8 core is a C-lattice equilibrium core of GE9B fuel.
In general, the calculated safety limit is dominated by two key
, parameters: (1) flatness of the bundle pin-by-pin power /R-factor distributions; and (2) flatness of the core bundle-by-bundle MCPR distributions. Greater flatness in either parameter yields more rods susceptible to boiling transition and thus a higher i
calculated SLMCPR., Hope Creek has a bundle R-factor distribution l more peaked than the generic GE9B equilibrium core. The core
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MCPR distributions for Hope Creek Cycle 8 is flatter than for the generic GE9B core.
The uncontrolled bundle pin-by-pin power distributions were compared between the Cycle 8 GE9B bundle which dominates the contribution to fuel pins in boiling transition and the GE9B bundle used in the generic SLMCPR analysis. For Hope Creek Cycle 8, the distribution of uncontrolled R-factors for the highest power rods in each bundle is not as flat as the bundle used in the generic analysis. For example, for Hope Creek Cycle 8, the
' bundles which contribute 4EEEEP of the pins undergoing boiling transition (out of the 0.100% of all pins in the core in boiling
> transitio was selected as this roughly corresponds to 0.01 in MCPR for GE9B fuel.
1 By keeping the limiting bundles uncontrolled, it is assured that the flattest possible pin-by-pin R-f actors are used in the SLMCPR calculation. By design, the R-factor distributions are optimized l
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I Document Control Desk LR-N97433 Attachment 1 LCR H97-05 for their uncontrolled state, and control blade insertion causes the distributions to become more peaked. Therefore, the most conservative approach is to perform the SLMCPR calculation where
- the " base" rod pattern places all the potentially limiting
! bundles in an uncontrolled state. The Hope Creek Cycle 8 SLMCPR anal sis has all of the bundles Ine i generic GE9B analysis has 'of the l core MCPR in an uncontroll state
! dauung>
uooe Creek Cycle 8 has for the generic GE9B core. These bundles near the core MCPR ar'e f ar more important to the determination of the SLMCPR that the bundles j of the core MCPR. The core MCPR distribution for Hope Creek l Cycle 8 is thus seen to be considerably flatter that for the l generic GE9B core. It is concluded that the greater flatness of i the Hope Creek Cycle 8 core MCPR distribution is enough to i overcome the greater flatness of the generic GE9B pin-by-pin R-j f actors and is the primary reason the calculated SLMCPR for the Hope Creek Cycle 8 core is 0.03 higher than the calculated SLMCPR for the generic GE9B equilibrium core. However, it should be pointed out that no specific sensitivity studies have been i performed for Hope Creek to quantify the relationship between
! SLMCPR and flatness as described in terms of percent of bundles l within a set delta CPR of the core MCPR or the number of pins l within a set delta R-factor of the limiting R-factor within a bundle.
A speci'ic single loop SLMCPR calculation was performed for Hope Creek Cycle 8. These calculations use the same procedure described above for cycle specific dual loop calculation, except that they apply the larger uncertainties specified by reference 4 j for single loop operation conditions. From the results of these j calculations, it was determined that the single loop operation t adder is 0.02 for Cycle 8 (for a single loop operation SLMCPR of 3
1.12). This is consistent with previous studies which have linearly correlated the single loop operation adder to the dual loop SLMCPR.
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. Document Control Desk LR-N97433 i Attachment 1 LCR H97-05 I I l l
Table 1: Comparison of Generic GE9B and Hope Creek Cycle 8 Cores Quantity, Description GE9B Hope i
Generic Creek Cycle 8 Nu:r.ber of bundles in core l 764 764*
l l Limiting cycle exposure point lMW t
i i
i 1
l Calculated Safety Limit MCPR 1.07 1.10 The Cycle 8 SLMCPR calculations are based upon a core consisting of 100% GE9B fuel: 236 fresh bundles (176 at 3.27%
l and 60 at 2.98% enriched), 232 once burnt bundles (88 at 3.25%
l and 144 at 3.24% enriched), 232 twice burnt bundles (88 at 3.25%
!' and 144 at 3.24% enriched) and 64 thrice burnt bundles (all at l 3.25% enriched).
l CONCLUSIONS:
l Based on all of the facts, observations and arguments presented above, it is appropriate to conclude that the calculated dual THE ATTACHMENTS TO THIS LETTER CONTAIN PROPRIETARY INFORMATION
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-Document Control DOsk
, Attcchment 1 LR-N97433 LCR H97-05 loop SLMCPR-value of-1.10 for the Hope Creek Cycle S core is reasonable. It is reasonable that this value ic 0.03 higher than the 1.07 value equilibrium core.
calculated for the generic C-lattice GE9B It is appropriate to conclude that the calculated Creek Cyclesingle 8 coreloop SLMCPR adder value of 0.02 for the Hope is reasonable.
It is reasonable that this value is 0.01 higher than the 0.01 value calculated for the generic C-lattice GE9B equilibrium core.
RE FERENCE S : -
- 1. General Electric letter to NRC, 10CFR Part 21, Reportable Condition, SLMCPR Evaluations, dated May 24, 1996.
- 2. Licensing Topical Report, General Electric BWR Thermal Analysis Basis (GETAB) : Data, Correlation and Design Application, NEDO-10958-A, January 19'7
- 3. General Electric Standard Application for Beactor Fuel, NEDE-2 4 011-P- A-13-US, August 1996.
A . General Electric Fuel Bundle Designs, NEDE-31152P, Reviston 6, April 1997.
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Document Centrol DO5k LR-H97433 Attachment 2 LCR H97-05 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE HPF-57 DOCKET NO. 50-354 SAFETY LIMIT MINIMUM CRITICAL POWER RATIO (SLMCPR) CHANGES 1
10CFR50.92 EVALUATION Public Service E.ectric & Gas (PSE&G) has concluded that the proposed changes to the Hope Creek Generating Station (HC)
Te7hnical Specifications do not involve a significant hazards
- censideration. In support of this determination, an evaluation of each of the three standards set forth in 10CFR50.92 is provided below.
REQUESTED CHANGE The proposed changes to the Hope Creek Technical Specifications contained in this submittal are being made to: 1) replace the 1.07 MCPR limit for two recirculation loop operation with a 1.08 MCPR limit for Hope Creek Cycl'e 7; and 2) replace the 1.08 MCPR limit for single recirculation loop operation with a 1.09 MCPR limit for Hope Creek Cycle 7 This Technical Specification change request also proposes to: 1) establish a 1.10 MCPR limit for two recirculation loop operation for Hope Creek Cycle 8; and
- 2) establish a 1.12 MCPR limit for single recirculation loop operation for Hope Creek Cycle 8.
BASIS
- 1. The proposed changos do not involve a significanc increase in che probability or consequences of an accident previously eval ua c ed.
The derivation of the revised SLMCPRs for Hope Creek for incorporation into the Technical Specifications, and its use to determine cycle-specific thermal limits, have been performed using NRC approved methods. Additionally, interim implementing procedures which incorporate cycle-specific parameters have been used which result in a more restrictive value for SLMCPR. These calculations do not change the method of operating the plant and have no effect on the probability of an accident initiating event or transient.
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Document Control Desk IJL-N97433 Attachment 2 LCR H97-05 There are no significant increases, in the consequences of an accident previously evaluated. The basis of the MCPR Safety Limit is to ensure that no mechanistic fuel damage is calculated to occur if the limit is not violated. The new SLMCPRs preserve the er.1 sting margin to transition boiling and the probability of fuel damage is not increased. Therefore, the proposed change does not involve an increase in the probability or consequences of an accident previously evaluated.
- 2. The prcposed change does not create che possibility of a new or d1!!erent kind of accident from any accident previously e val ua t ed.
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The proposed changes contained in this submittal result from an analysis of the Cycle 7 and Cycle 9 core reloads using the same fuel types a: previous cycle.=, These changes do not involve any new method for operating the facility and do not involve any facility modifications. No new initiating events or transients result from these changes. Therefore, the proposed Technical Specification changes do not create the possibility of a new or different kind of accident, from any accident previously evaluated. ,
- 3. The prcposed change does not involve a significant reduction in a margin of safecy.
The margin of safety as defined in the Technical Specification bases will remain the same. The new SLMCPRs are 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 been 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 violated, thereby p eserving the fuel cladding integrity. Therefore, the proposed Technical Specification changes do not involve a reduction in a margin of safety.
CONCLUSION Based on the above, PSE&G has determined that the proposed changes do not involve a significant hazards consideration.
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