ML20063C406

Application for Amend to License DPR-29,proposing Tech Spec Changes to Implement 10CFR50.59 Reload Licensing W/Odyn Transient Analyses & Extended Exposure MAPLHGR Limits. Significant Changes Discussed.Class III Amend Fee Encl

ML20063C406
Person / Time
Site:	Dresden, Quad Cities, 05000000
Issue date:	08/19/1982
From:	Rausch T COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML20063C409	List: ML20063C406 ML20063C433 ML20063C487
References
4658N, NUDOCS 8208270250
Download: ML20063C406 (5)

Text

. , - _

t l CN Commonwealth Edison 4

) one Fast National Plaza. Chicago. Ilknois C~ Address Reply to: Post Office Box 767 Chicago Illinois 60690 Augus t 19, 1982 i

Mr. Harold R. Denton, Director Of fice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Quad Cities Station Unit 1 Propose d Amendment to Appendix A Technical Specifications to Facility Operating License DPR-29 to Implement 10 CFR 50.59 Reload Licensing with ODYN Transien t Analyses and Ex tended Exposure MAPLHGR Limits NRC Docket No. 50-254 References (a): Letter, T. A. Ippolito to J. S. Abel dated Decembe r 5, 1980 (QC -1 Am . 61)

(b): NEDO-24146A " Loss-o f-Coolant Accident Analysis Report for Dresden Units 2, 3 and Quad Cities 1, 2 Nuclear Power Stations" Revision 1, April, 1979 as subsequently modified by Erratta and Ad denda No s.1, 2, 3, 4, 5, 6, and 7 ( E& A No . 7 is attached).

, (c): NEDO-24154 a nd NEDE-2415 4-P, Volumes I, I I, and III, " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors,"

October, 1978.

(d): Letter, R. H. Buchholz ( GE) to P. S. Check,

" Response to NRC Request for Information on ODYN Computer Model," September 5, 1980.

Cygy, (e): Le t te r , R . H. Buchholz (GE) to P.S. Ch e c k ,

"ODYN Adjustment Methods for Determination of Operating Limits," January 19, 1981. g,f0 I.D f '

i

( f): W Letter, R. F. Ja necek to H. R. Denton dated May 12, 19 81.

OO YO (g): Letter, E. D. Swartz to H. R. Denton da t e d July 2 7, 19 81.

Dear Mr. Denton:

Pursuant to 10 CFR 50.59, Commonwealth Edison proposes to amend Appendix A, Technical Specificatlons, to Facility Operating License DPR-29 to support the review o f future reloads for Quad Cities Unit 1 by Commonwealth Edison in accordance with the provisions o f 10 CFR 50.59.

8208270250 820819 PDR ADOCK 05000254 l P PDR

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H. R. Denton Augus t 19, 1982 The preparatory changes approved previously in Reference (a) do not sufficiently bound the impending Quad Cities Unit 1 Reload 6 Cycle 7 reload primarily because this will be the first time the unit is: analyzed with the ODYN transient code, and because extended exposure MAPLHGR analyses have been completed.

The proposed changes in At tachment 1 have received On-Site and Off-Site review and approval. The significant changes are discussed below, and are very .similar to those approved in Quad Cities Unit 2 Amendment No . 69 ( T. Ippolito letter to L. DelGeorge dated December 23, 1981).

MAPLHGR Limits NEDO 24146A (Reference (b)) contains the previously approved ECCS analysis for Dresden Units 2 and 3 and Quad Cities Units 1 and 2 and a

continues to serve as the basis for the generation of MAPLHGR limits for new fuel types.

The attached proposed Technical Specification changes include revisions to the MAPLHGR limit curves for fuel types 80RB265L, P8DR8265L, P80RB265H, and P80RB282. The revised curves extend the MAPLHGR limits

from the- previous maximum planar average exposure o f 30,000 MWD /ST out to a planar average exposure o f 40,000 MWD /ST, and separate limits for prepressurized fuel from non-pressurized fuel for fuel type 8DRB265L.

4 The maximum nodal exposure expected for the remaining lifetime of these fuel types is less than 40,000 MWD /ST.

DELETION OF 7x7 FUEL LIMITS MCPR, MAPLH GR , and LHGR operating limits for 7x7 fuel and the mixed oxide fuel (type EEIC-Pu) have all been removed as there will be no 7x7 fuel or mixed oxide fuel in the Quad Cities 1 Cycle 7 core. As a result of these deletions, the Figure 3.5-1 sheet numbers have been revised.

PRESSURE SAFETY LIMIT CHANGES DUE TO ATWS RPT The NRC required installation and implementation o f Recircula-tion Pump Trip (RPT) for ATWS mitigation as o f January 1, 19 81. Althou gh it reduces peak pressures for transients without scram, it carries a side effect of increasing the peak pressures for severe pressurization events eith scram (such as LR w/o BP or MSIV closure w/o valve position trip).

I On the positive side, pressurization events whlch exceed the RPT set point (1250 psig) can reach high steam dome pressures without exceeding the peak vessel or coolant system pressure criteria as a result of the lower reactor pressure drop which occurs without forced recirculation flow. Without RPT maximum pressure differences from the steam dome to l the bottom of the vessel were less than 30 psi. With RPT the total reactor A P is reduced to less than 20 psi.

a t

H. R. Den ton August 19, 1982 The assumption in the current bases for the pressure safety limit is a 50 psi A P (i.e.1375 psi -1325 psi) which is conservative in either case. The proposed change retains adequate conservatism by resetting the safety limit at only 1345 psig as measured in the steam dome. The assumed pressure difference is still 30 psig to the bottom o f the vessel which will assure compliance with the ASME code criteria of 110% of vessel design pressure (i.e.110% x 1250 = 1375 psig). Since the vessel peak pressure (bottom) is specifically calculated for each reload for the postulated multiple failure MSIV closure event (no valve position trip scram ana no Electromatic Relie f Valve Flow), the proposed change does not af fect our ability to identify potential problems with ASME compliance.

Wording changes in the bases have also been incorporated which clarify that compliance of the peak vessel pressure with the ASME criteria also assures compliance of the primary system piping pressures with the USASI criteria for the limiting point (i.e. less than 1410 psig at the lowest point in the recirc. suction line). These changes were recommended by GE due to the false implication in the current bases tha t all points in the primary system must remain less than the ASME criteria for the vessel (1375 psig). .

ODYN THANSIENT CODE IMPLEMENTATION The ODYN transient analysis computer code is used for analyzing rapid pressurization events in a more sophisticated manner than its predecessor, the REDY code. Re ference (c) is the three volume generic topical report which describes the model and its qualification.

Re ference (d) contains responses to NRC questions but also forms the primary reference for the implementation procedure, i.e. adjustment of ODYN MCPR results to account for statistical treatment of four parameters:

l

a. initial core thermal power

b. CRD scram times i

c. model uncertainty

d. regressional fitting uncertainty This statistical approach (" Option B") was negotiated with the staf f l during the first 9 months of 1980 and is intended to establish a "95/95" i basis for licensing ODYN. Tha t i s , the MCPR operating limit should provide a 95% probability with 95% confidence that the limiting pressurization event will not cause MCPR to fall below the fuel cladding integrity sa fety limit. In order to accomplish this, a statistically based scram time distribution which is faster than those in the current Technical Specifications was applied for each of several plant groupings of similar design in order to define generic Statistical Adjustment Factors (SAFs) which can be applied to plant specific results. Th e S AFs also incorporate the statistical treatment of power level, model, and fitting uncertainties. The net result for BWR2 and 3's is a SAF o f I

' o0.006 which is added to the ODYN calculated A CPR/ICPR . Th e ne w ICPR i s then calculated using the Reference (f) equation of:

O 4

H. R. Denton Augus t 19, 1 2 SL I CPRne w = 1 *&CPRg + .006 '

ICPR C

• where SL=MCPR Safety Limit (1.07)

CPRc = ODYN calculated transient & CPR (unadjusted)

ICPRc = A CPRc + SL To assure L.1d demonstrate consistency of operations with the assumed scram time distribution in the calculation of the SAFs, a " scram time conformance procedure" is now required which basically makes the MCPR operating limit a function of scram times as measured during the normal surveillance. Specifically, the overall average of all 20%

insertion scram time data measured to date in the current cycle (Tave) must be evaluated with respect to the 5% significance level criteria for the distribution (T )Bassumed in deriving the generic SAF. If the running average exceedstb a MCPR penalty is required.

The MCPR penalty is applied in the form o f a linearly increasing limit between the Option 8 value at t b to a more conservative NRC-determined value (" Option A" MCPR limit) a t t ave, which is the CRD surveillance limit for 20% insertion from specification 3.3C.2.

The Option A limit is also defined in References (d) and (e) and is simply equal to ICPRc multiplied by 1.044 (i.e. a 4.4% penalty on the unadjusted ODYN results).

The scram time dependence o f the MCPR limit is reflected in the proposed changes to Technical Specifications and bases sections

3. 3. c/ 4. 3. c a nd 3. 5. k . The Option 8 MCPR limits were chosen as 1.37 for 8 x 8 and 8 x 8R and 1.39 for P8 x 8R for these preparatory changes which bounds the Quad Cities 1 Cycle 7 reload transient analyses results. If these limits are not suf ficient for a future reload, revised preparatory Technical Specifications will be required.

The proposed form of the scram time dependent MCPR limits is an attempt to simplify the actual Technical Specification implementation o f the conformance procedure. General Electric originally suggested either an explicit version (full equations and definitions in the LCO text) or a graphical version. Both methods are extremely complex and would have presented significant challenges in plant implementation and operator training. The proposed approach, while still complicated relative to previous MCPR specifications, o ffers significant advantages over other alternatives while incorporating small conservatisms which should not impact plant operation.

4 H. R. Denton Augus t 19, 1982 The conservatisms include:

a. The assumption of maximum scram timing frequency allowed by Specification 4.3.C.2 (that is, a full core data set at BOC and half core data sets every 16 weeks thereaf ter) .

b. the assumption of an operating cycle length of 24 months (excluding refueling) which is longer than what is currently considered technically feasible without excessive coastdown and associated economic constraints.

c. a conservative choice of the nearest RPIS switch (dropout of pos. 39) in the selection of the mean (jk ) and standard deviation ( r) associated with calculating the 5% significance level criteria ( T B),

d. conservative rounding o f both 1EB and IA values, and e, use of the rounced down TB and IA values in calculation of the slope and intercept values for the linear MCPR penalty between TB and TA.

Pursuant to 10 CFR 120, Commonwealth Edison has determined that this proposed amendment is Class III. As such, a fee remittance in the amount o f $4,000 has been enclosed.

Please address any questions you may have concerning this matter to this of fice.

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

Very truly yours, f ke:"

Thoma s J. Rausch Nuclear Licensing Administrator 1m Attachment 1. Proposed Changes to DPR-29

2. Errata and Addenda Shee t No . 7 to NEDO 24146A cc: RIII Inspector - Quad Cities SUBSCRIBED and SWORN to before me this /N, day o f 81uve 4 , 1981 h4 2. . #+

Notary Public 4658N