ML20248L802

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Proposed Tech Specs Section 2.1.2,re Thermal Pressure,High Pressure & High Flow
ML20248L802
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Site: Fermi DTE Energy icon.png
Issue date: 06/05/1998
From:
DETROIT EDISON CO.
To:
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ML20036E415 List:
References
NUDOCS 9806110368
Download: ML20248L802 (9)
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NRC-98-0067

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ENCLOSURE 3 j

FERMI 2 NRC DOCKET NO. 50-341 OPERATING LICENSE NO. NPF-43 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:

THERMAL POWER, High Pressure and High Flow Attached is a mark-up of the existing Technical Specifications (TSs), indicating tne

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proposed changes (Part 1) and a typed version of the TSs incorporating the proposed changes with a list ofincluded pages (part 2).

9806110368 900605 PDR ADOCK 05000341 P

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. to WRC-98-0067 Page 2 ENCLOSURE 3 - PART I PROPOSED TECHNICAL SPECIFICATION MARKED UP PAGES INCLUDED PAGE(S):

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l 2.0 SAFFTY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER. Low Pressure or low Flow l

2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than l

10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION.

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the

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requirements of Specification 6.7.1.

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THERMAL POWER. Hiah Pressure and Hiah Flow

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2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than the Safety Limit MCPR of or two recirculation loop operation and shall not I

l be less than the Safety Limit MCPR of for single loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

g.g APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

l fl l *l3 With MCPR less than the Safety Limit MCPR of 4-09 forf two recirculation loop operation or less than the Safety Limit MCPR of M.for single loop operation I

and with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

1 REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

FERMI - UNIT 2 2-1 AmendmentNo.M,D,/)ijf 1

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NRC-98-0067 ~

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ENCLOSURE 3 - PART 2 l

PROPOSED TECHNICAL SPECIFICATION REVISED PAGES l

INCLUDED PAGE(Sh 2-1 l-I I

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g,0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITI THERMAL POWER. Low Pressure or Loy Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel-steam dome pressure less than 785 psig or core flow less than 10% of rated flow, i

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

THERMAL POWER. Hiah Pressure and Hiah Flow

  • l 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than the Safety Limit MCPR of-1.11 for two recirculation loop operation and shall not l

l be less than the Safety limit MCPR of 1.13 for single loop operation with the i

reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

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' APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTIOM:

With MCPR less than the Safety Limit MCPR of 1.11 for two recirculation loop opers' ion or less than the Safety Limit MCPR of 1.13 for single loop operation and with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 i

L hours and comply with the requirements of Specification 6.7.1.

I REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The. reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

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APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3 and 4.

I ACTION:-

With the reactor coolant systn pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

operation only.

p FERMI - UNIT 2-2-1 Amendment No ff, JJ, J##

d to NRC-98-0067 Page1 ATTACHMENT 3 GENERAL ELECTRIC NUCLEAR ENERGY DUi.'UMENT ENTITLED

" ADDITIONAL INFORMATION REGARDING TIIE 1.11 CYCLE SPECIFIC SLMCPR FOR FERMI-2 CYCLE 7" (NON-PROPRIETARY VE.RSION)

GE COMPANY I

l Attachment AdditionalInformation Regarding the 1.11 May 8,1998 Cycle Specific SLMCPR for Fermi-2 Cycle 7 References

[l]

General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, NEDO-10958-A, January 1977.

(2)

General Electric Standard Applicationfor Reactor Fuel (GESTAR 11), NEDE-24011-P-A-11, November 1995.

[3]

Smith, M.A. (GE) letter to NRC Document Control Desk,10CFR Part 21, Reportable Condition, Safety Limit AfCPR, letter # 96-04NRC. DOC (MFN 074-96), May 24,1996.

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[4]

General Electric Standard Applicationfor Reactor Fuel (GESTAR 11), NEDE-240 l 1 -P-A-12, June 1996.

[5]

Matthews, David B. (NRC) letter to Ralph J. Reda (GE), Proposed General Electric Revision 12 to GESTAR11, August 5,1996.

[6]

General Electric Standard Applicationfor Reactor Fuel (GESTAR 11), NEDE-240 l 1-P-A-13, August 1996.

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Reda, R.J. (GE) letter to NRC Document Control Desk with attention to T.E. Collins, Acting Branch Chief of Reactor Systems Branch: Proposed Amendment 25 to GE Licensing Topical Report NEDE-21011-P-A (GESTAR 11) on Cycle-Specific Safety Limit AfCPR; December 13,1996.

[8]

Reda, R.J, (GE) letter to NRC Document Control Desk with attention to T.E. Collins, Acting Branch Chief of Reactor Systems Branch: GESTAR Amendment 25 Supporting Information:

December 13,1996.

[9}

GeneralElectric FuelBundle Designs, NEDE-31152-P, Revision 5, June 1996.

i l10) Afethodolog and Uncertaintiesfor Safety Limit AfCPR Evaluations, NEDC-32601 P, Class 1II, December 1996. (Submitted for NRC review by letter from R.J. Reda (GE) to NRC Document Contvl Desk dated December 13,1996.)

ll i1 ' R-Factor Calculation Afethodfor Gell, GE12 and GE13 Fuel, NEDC-32505P, November 1995.

Background

The AFC (1974) approved methodology for calculating the Safety Limit Minimum Critical Power Ratio (SLMCPR) is documented in GETAB[Il which is specifically referenced by GESTAR. II.

Prior to 1996, GESTAR 11[2] stipulated that the SLMCPR analysis for a new fuel design be i

performed for a large high power density plant assuming a bounding equilibrium core. The Gell product line generic SLMCPR value of 1.07 was determined according to this specification.

j in March 1996 it was discovered that the SLMCPR calculated on a generic basis with the GE raethodology may be non-conservative when applied to some actual core and fuel designs. A 10CFR Part 21, Reportable Condition [3] documenting the problem and the earlier communications and mectmgs with the NRC as well as the NRC inspection No. 99900003/96-01 in Wilmington May 6-10 page1of3 L_________---______

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GE COMPANY Attachment Additional Information Regarding the 1.11 May 8,1998 Cycle Specific SLMCPR for Fermi-2 Cycle 7 was filed on May 24,1996. The NRC found in their inspection that GE " failed to recalculate or reconfirm the applicability of the generically determined SLMCPR to new fuel bundle designs".

Revision 12 to GESTAR 11[4] was submitted in response to this finding in order to clarify how the calculated SLMCPR would be reconfirmed on a plant / cycle specific basis. The NRC staff requested [5] "that General Electric withdraw this revision and submit an amendment incorporating full information on the SLMCPR analysis processes". In the same letter [5] the NRC staff stated that SLMCPR values "should be confirmed for each plant / core-specific application". In response to this NRC request, GE issued Revision 13 ta GESTAR 11[6] without the description of how the plant / cycle specific SLMCPR methodology would be applied to reconfirm the safety limit. Instead this information was submitted [7] as requested by the NRC as the " Proposed Amendment 25.".

Pending NRC approval of the proposed Amendment 25 to GESTAR II, the NRC staff has been reviewing plant / cycle specific Safety Limits individually for each cycle of each plant. The purpose of this document is to provide the NRC with additional information to support their review of how plant / cycle specific SLMCPR analyses using the uncertainties defined in Reference [9] were used to confirm the calculated SLMCPR value for Fermi-2 Cycle 7.

Comparison of Fermi-2 Cycle 7 SLMCPR versus the Generic gel 1 Value Table 1 summarizes the relevant input parameters and results of the SLMCPR determination for both the generic GElI core and the Fermi-2 Cycle 7 core. Both generic and plant / cycle specific evaluations are performed using the methods described in GETABill. The evaluations yield different calculated SLMCPR values because the inputs that are used are different. The quantities that have been shown to have some impact on the determination of the safety limit MCPR (SLMCPR) are provided. Much of this information is redundant but is provided in this case because it has been provided previously to the NRC to assist them in understanding the differences between plant / cycle specific SLMCPR evaluations and the generic values calculated previously by fuel product line.

l In comparing the generic Gell and Fermi-2 Cycle 7 SLMCPR values it is important to note that the Fermi-2 Cycle 7 core is approaching a Gell equilibrium core. The freshest fuel is the latest batch of gel 1 that comprises a fraction of the bundles in the core. Also, this fresh batch of GE1I has the l

highest enrichment, as compared to the core average enrichment, as shown in Table 1. By way of comparison, the generic Gell equilibrium core has different batch and core average enrichments.

Higher enrichment in the fresh fuel for the Fermi-2 Cycle 7 core (compared to the rest of the core) produces higher power in the fresh bundles relative to the rest of the core. These enrichment differences r~sult in the Gell fresh fuel producing a higher relative share of the number of fuel rods l

calculated to be susceptible to boiling transition (NRSBT).

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-By all important measures the Fermi-2 Cycle 7 core MCPR distribution is evaluated as being flatter than the core MCPR distribution used in the Gell generic analysis and bundle R-factor distributions are also flatter.

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GE COMPANY Attachment AdditionalInformation Regarding the 1.11 May 8,1998 Cycle Specific SLMCPR for Fermi-2 Cycle 7 The uncontrolled bundle pin-by-pin power distributions were compared between the Fermi-2 Cycle 7 bundles and the GElI bundle used in the generic SLMCPR analysis. Pin-by-pin power distributions are characterized. terms of R-factors using the methodology defined in Reference [11). For the Fermi-2 Cycle 7 ba dies, there is a slightly flatter distribution of uncontrolled R-factors for the highest power rods ic each bundle, which in the calculation are the rods most likely to be susceptible to boiling transition. This fact is suggested by a graphical comparison of the bundle R-factor distributions but is difficult to quantify graphically since the relative flatnesses are similar and the rods that have an R-factor closer to the R-factor for the lead rod are statistically worth much more than those that have R-factors that are further away.

Thus this supports the conclusion that the higher SLMCPR value for Fermi-2 Cycle 7 is at least in part due to the flatter bundle R-factors relative to those used for the generic gel 1 SLMCPR evaluation.

Table 1. Comparison of gel 1 Generic and Fermi-2 Cycle 7 Core and Bundle Quantities that Impact the SLMCPR Summary The calculated nominal 1.11 Monte Carlo SLMCPR for Fermi-2 Cycle 7 is consistent with what one would expect. Various quantities have been used over the last year to compare quantities that impact the calculated SLMCPR value. These other quantities have been provided to the NRC previously for other plant / cycle specific analyses using a format such as that given in Table 1. These other quantities have also been compared for this core / cycle. The key parameters in Table I support the conclusions that the Fermi-2 Cycle 7 core / cycle has a flatter core MCPR distribution and flatter in-bundle power distributions than what was used to perform the gel 1 generic SLMCPR evaluation.

Both the core MCPR distribution and the bundle R-factor distribution contribute to the higher calculated SLMCPR relative to the generic gel 1 SLMCPR evaluation.

Based on all of the facts, observations and arguments presented above, it is concluded that the calculated SLMCPR value of 1.11 for the Fermi-2 Cycle 7 core is appropriate. It is reasonable that this value is 0.04 higher than the 1.07 value calculated for the generic gel 1 equilibrium core.

For single loop operations (SLO) the safety limit MCPR is 0.02 greater than the two loop value.

1 Prepared by:

Verified by:

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R. H. Szilard, Ph.D.

K. M. Fawc'tt

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e Technical Project Manager Engineer Fermi-2 Project Nuclear Safety & Analysis l

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