ML20198B912

Application for Amend to License NPF-11,revising Tech Specs in Support of First Reload.Ge Rev 0 to 23A1843, Supplemental Reload Licensing... & Proprietary Figures Also Encl.Figures Withheld (Ref 10CFR2.790).Fee Paid

ML20198B912
Person / Time
Site:	LaSalle
Issue date:	10/22/1985
From:	Massin H COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML19344C148	List: ML20198B912 ML20198B917 ML20198B931
References
0797K, 797K, NUDOCS 8511070359
Download: ML20198B912 (7)

Text

r

^

I- / N Commonwealth Edison

( ) One First Nation'l PI z% Chicago, Ilknois

( C ] Addr;ss Reply 1: Post Offica Box 767

\ ,/ Chicago Illinois 60690 t.

October 22, 1985 Mr. Harold R. Denton, Director l Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

LaSalle County Station Unit 1 Proposed Amendment to Technical Specification for Facility Operating License tPF-ll - Reload Licensing Package for Cycle 2 l

NRC Docket Nos. 50-373 1

l References (a): J. F. Quirk (CE) to 0. D. Parr (NRC),

" General Electric Licensing Topical Report NEDE-240ll-P-A, Generic Reload j Fuel Application," 9/11/78.

(b): GE Document, NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel," (CESTAR II).

(c): NRC Memorandum, MFN-061-85, C. O. Thomas (NRC) to H. C. Pfefferlen (NRC), " Acceptance for Referencing of Licensing Topical Report NEDE-240ll, Rev. 6, Amendment 8, ' Thermal Hydraulic Stability Amendment to GESTAR II',"

dated April 24, 1985.

(d): NRC Memorandum, L. S. Rubenstein (NRC) to G. C. Lainas (NRC), " Changes in CE Analysis of the Control Rod Drop Accident for Plant Reloads (TACS-48058)," dated February 15, 1983.

Dear Mr. Denton:

Pursuant to 10 CFR 50.59, Commonwealth Edison proposes to amend Appendix A, Technical Specification, to Facility Operating License NPF-ll.

These changes are being submitted for your staff's review and approval and are in support of the first reload for LaSalle Unit 1. Startup for Cycle 2 f is currently scheduled for March,1986.

gO Attachment A provides background and discussion. The proposed changes are enclosed in Attachment B. The attached change has received both On-Site and Off-Site review and approval. We have reviewed this amendment request and find that no significant hazards consideration exists. Our y review is documented in Attachment C. Attachment 0 is the GE reload licen-o sing submittal. Attachment E is a non-proprietary version of licensing O information to support the new bundle enrichment of the reload fuel for k Cycle 2. The three proprietary pages are submitted under se cover

&!l@k & stL %A%g

H. R. Denton October 22, 1985 (Attachment F), with the request that they be treated as proprietary and withheld from public disclosure in accordance with the provisions of 10 CFR .

2.790, under the GE affida/it of proprietary information submitted with the i original document (Reference (a)).

Commonwealth Edison is notifying the State of Illinois of our request for this amendment by transmitting a copy of this letter and its attachments to the designated State Official.

In accordance with the requirements of 10 CFR 50.170, a fee remittance in the amount of $150.00 is enclosed.

Please direct any questions you may have concerning this matter to this office.

Three (3) signed originals and thirty-seven (37) copies of this transmittal and its attachments are provided for your use.

Very truly yours, H. L. Massin Nuclear Licensing Administrator 1m Attachments A: Background and Discussion 8: Technical Specification Change to NPF-11 C: Evaluation of Significant Hazards Consideration D: GE Document, 23A1843, " Supplemental Reload Lirensing Submittal for LaSalle County Station, Unit 1, Reload 1 (Cycle 2)," dated June 1985 (5 pgs. Add. Info. Attached)

E: GE Letter, REP: 85-094, R. E. Parr (GE) to J. L.

g Anderson (CECO), "LaSalle Fuel Bundle OP8CRff99L,"

dated May 17, 1985 (Non-Proprietary Version)

F: Proprietary Version of Attachment E cc: Region III Inspector - LSCS A. Bournia - NRR M. Parker - State of Ill SUBSCRIBE 0 AND SWORt to befor r* /L day of , , F{phis ju , 1985 r_ LA h 04)1 Notary Public j 0797K

ATTACHMENT A TECHNICAL SPECIFICATION CHANGE REQUEST LASALLE COUNTY STATION UNIT 1 CYCLE 2 RELOAD LICENSING SUBMITTAL l

BACKGROUND LaSalle Unit 1, Cycle 2 will utilize 232 GP8CR8299L fuel bundles.

The fuel type is prepressurized barrier fuel. Information on the Cycle 2 reload may be found in the " Supplemental Reluad Licensing Submittal for LaSalle County Station, Unit 1, Reload 1 (Cycle 2)," 23A1843, and in the Additional Information documents (Attachment D).

Fuel type BP8CRB299L is not yet included in the General Electric Standard Application for Reactor Fuel (Reference (b)) and therefore must be l

licensed. This fuel type will be incorporated in GESTAR as part of the next amendment. GE has provided a letter (Attachment E) including the information

, required to license the fuel type. As stated in the letter the new fuel

" type has been analyzed with approved methods and meets the approved limits of GESTAR. The new fuel type presents no unreviewed safety questions because the BP8X8R bundle design has been approved by the NRC, and licensing of new bundle enrichments has been treated as non-safety related change to GESTAR.

DISCUSSION

1. MCPR Safety Limit The FCPR fuel cladding integrity safety limit for Cycle 2 is 1.07. This is an increase over the 1.06 allowed for the initial core. See Table S.2-3a of Reference (b). The safety limit is smaller for initial cores because the uncertainties in TIP readings and the R Factor are smaller.

2. Limiting MCPR Transient The Cycle 2 MCPR operating limits required to preclude violation of the fuel cladding integrity safety limit are 1.24 for 8X8R fuel and 1.26 for BP8X8R fuel. These values are based on the Feedwater Controller Failure (FWCF) event analyzed with the 00YN Option B approach. This value is an increase of 0.02 over the initial cycle and requires a Tech Spec change to Figure 3.2.3-1.

The slope change in Figure 3.2.3-1 indicates the change in limiting transients. Fort less than 0.755 the limiting transient is the FnCF event. At 7 equais 0.736 the FCPR limit for Rod Withdrawal Error (RWE) is the same as the limit for FWCF. For T greater than 0.755 the limiting transient is the Load Reject without Bypass (LR w/o OP) event.

Other MCPR related Tech Spec (3/4.2.3) changes are:

The replacement of the present Kr curve with a revised curve, Figure 3.2.3-2. The revised curve is based on a core power of 3323 MWth and a core flow of 108.5 Mlb/hr. The original curve was a generic curve.

The deletion of the EOC-RPT inoperable provision in the specification.

The analysis was not justified for the second cycle but may be included in future cycles.

3. Loss of Feedwater Heating o The Loss of Feedwater Heating (LOFnH) event was analyzed using the GE BwR Simulator Code rather than the REDY code. As stated in GESTAR (Reference (b)) Section S.2.2, slower core-wide transients such as LOFwH can be analyzed by using either of the codes. The simulator code results, while still conservative, are more realistic than the RE0Y results.

A feedwater temperature change of 1450F was assumed for the LOFWH event. A 1450F change will bound the temperature change of the most probable LOFhH events.

4. Compliance to ASME Pressure Vessel Code The results of the LIC2 analyses for the postulated MSIV closure with flux scram provided in Attachment A indicate that the peak steamdome pressure will be 1238 psig and the peak vessel pressure will be 1269 psig. These values are less than the steam dome pressure safety limit of 1325 psig from the Tech Specs and the ASME vessel overpressurization limit of 1375 psig (110% of design pressure). Because the calculated values are less than the limits the pressure response is acceptable.

5. Rod Withdrawal Error The Rod Withdrawal Error (RwE) has been analyzed on a plant / cycle specific basis. The RBM rod block selected setpoint of 107 gives a b CPR of 0.19 for both the BP8X6R and 8X8R fuel. AdJing this ACPR to the safety limit of 1.07 yields an event LCO of 1.26. This is equal to the value for the FWCF event, so the RWE and the FwCF are the bounding events.

6. Fuel Loading Error Event No Fuel Loading Error analysis is required for LIC2. Neither mislocated nor misoriented bundle events are analyzed for BWR 5 reloads.

The mislocated bundle accident is only performed for initial cores.

Data from past reloads indicate that the probability of mislocating a fuel bundle so that the CPR violates the safety limit is sufficiently small that plant specific analyses are unnecessary. The NRC has given interim approval to this approach. See GESTAR (Reference (b)) section S .2 . 5. 4 .1.

The misoriented bundle accident is not analyzed for C-lattice cores,

. such as LaSalle, because the misorientation causes an insignificant CPR i change. This is due to the symmetry of the fuel bundles. Proper orientation during core loading is also readily verified visually. For discussion see GESTAR (Reference (b)) section S.2.5.4.2.

7. Stability Analysis i

The LIC2 decay ratio at the intersection of the natural circulation line and the 105 per cent rod line is 0.60. Since existing Technical Specifications do not allow continued operation in natural circulation, i combinations of low flow and high power sufficient to produce high decay I ratios are not permitted.

The cycle specific decay ratio of 0.60 is less than the NRC upper limit of 0.80 for plants with no stability monitoring technical specifications (Reference (c)). Since the decay ratio is within limits, the stability of the reload is acceptable. Although the cycle specific analysis is sufficient to demonstrate adequate stability, proposed Tech Spec changes for stability monitoring have been included to address NRC concerns in this area.

Proposed Tech Spec changes for single loop operation are also included.

Single loop operation was allowed during the first cycle but must be reapproved because it was only approved for the first cycla.

8. Loss of Coolant Accident The MAPLHGR limits for the new fuel type BP8CRB299L are included in the CE licensing document (Attachment D). The curve for these limits will be added to the LaSalle County Station Unit 1 Technical Specifications.

9. Rod Drop Accident The Rod Drop Accident (RDA) event has been statistically analyzed on a generic basis and is no longer analyzed on a plant cycle specific basis. The generic analysis provides assurance that the 280 cal / gram enthalpy oeposition limit will not be violated. The highest deposition of enthalpy calculated was 158 cal / gram. This provides confidence on 95/95 level that the Technical Specification limit will not be violated in the unlikely event of the postulated Design Basis RDA. The generic RWE analysis has been approved by the NRC (Reference (d)).

TECHNICAL SPECIFICATION CHANGES The following Technical Specification changes will support operation of LaSalle County Station Unit It1 during Cycle 2:

Technical Specification 2.1, Safety Limits.

The MCPfs fuel cladding integrity safety limit was changed from 1.06 to 1.07 for two recirculation loop operation, end from 1.07 to 1.08 for single recirculation loop operation.

Safety Limits Bases.

Bases Tables 82.1.2-1, B2.1.2-2, 62.1.2-3, and 82.1.2-4 were changed to reflect the reload core inputs to the statistical model which determines the MCPR safety limit (GETAB-General Electric Thermal Analysis Basis). The reload values were obtained from CESTAR (Table S.2-1, S.2-2, S.2-2c and Figure S.2-la).

The thermal power and core flow inputs to the model varied from rated unit conditions (thermal power was 3293 MWth rather than 3323 MWth, and core flow was 102.5 M1b/hr rather than 108.5 M1b/hr). This had insignificant imp 6ct on the PCPR safety limit because the analysis performed is considered bounding for a 3323 MWth plant.

Technical Specification 3/4.2.1, Average Planar Lineat Heat Generation Rate.

Figure 3.2.1-2, which contains the MAPLHGR vs Exposure curve for the reload fuel type DP8CRB299L, was added to the specification.

Figures 3.2.1-1 and 3.2.12, MAPLHCR vs Exposure Curves.

The fuel type numbers were changed in the legend and title on Figure 3.2.1-1 to reflect the correct initial core fuel types.

Figure 3.2.12 contains the MAPLHGR vs Exposure curve for the reload fuel type 6PSCRB299L. A separate graph was used for the reload fuel curve in order to avoid confusion with initial fuel curves.

Technical Specification 3/4.2.3, Minimum Critical Power Ratio.

The EOC-RPT . inoperable provision in this specification has been deleted. The EOC-PPT inoperable analysis was not justified for the second cycle but may be included in future cycles.

Figure 3.2.3-1, MCPR vs.T at Rated Flow.

Present curve has been replaced by a curve which reflects the change in the limiting transients for cycle 2. For 7 less than 0.755 the limiting transient is the Feedwater Controller Failure (FwCF) event. At 7 equals 0.736 the MCPR limit for Rod Withdrawal Error (RwE) is the same as the limit for FhCF. For 7 greater than 0.755 the limiting transient is the Load Reject Without Bypass (LR w/o OP) event.

f t Figure 3.2.3-2, Kr Factor.

Present curve has been replaced by a revised curve which is based on a core power of 3323 MWth and a core flow of 108.5 M1b/hr. The original curve was a generic curve.

Technical Specification 3.3.4.2, End-of-Cycle Recirculation Pump Trip System Instrumentation.

Action statements d end e were revised since the E0C-RPT inoperable action statement in specification 3.2.3 was deleted. Power reduction requirements (less than 30% of rated thelmal power within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) were inserted in the action statements in the event restoration of inoperable trip systems cannot be accomplished within specified time requirements. l Technical Specification 3/4.4.1, Recirculation System.

Single loop operation action statements addressing NRC concerns on stability have been deleted since a stability specification is being inserted with this amendment.

The MCPR safety limit for single loop operation was changed from 1.07 to 1.08.

Technical Specification 3/4.4.1.5, Thermal Hyoraulic Stability.

Thermal hydraulic stability specification was aoded to address NRC concerns in this area.

Recirculation System bases.

The Bases for specification 3/4.4.1 has been changed to reflect the fact that single recirculation loop operation will be allowed for future cycles.

A paragraph was added to the bases for specification 3/4.4.1.5 (Thermal Hydraulic Stability).

0797K