ML17319B551

Second Suppl to 820407 Application for Cycle 4 Reload & Uprate License Amend.Further Changes to App a Tech Specs Necessary to Allow Cycle 4 Operation W/Exxon Nuclear Co Fuel at 3,411 Mwt

ML17319B551
Person / Time
Site:	Cook
Issue date:	09/30/1982
From:	Hunter R INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	Harold Denton Office of Nuclear Reactor Regulation
References
AEP:NRC:0637E, AEP:NRC:637E, NUDOCS 8210050211
Download: ML17319B551 (17)

Text

REGULATORY

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ACCESSION NBR;8210050211 DOC ~ DATE: 82/09/30 NOTARIZED: NO DOCKET FACIL:50 BYNAME 316 Donald C+ Cook Nuclear Power Plant~ Unit 2q Indiana 8 05000316 AUTH AUTHOR AFFIL I ATION HUNTERtR ~ St Indiana L Michigan Electric Co ~

REC IP ~ NAME RECIPIENT AFFILIATION DENTON~H ~ Ro Office of Nuclear Reactor Regu)ationi Director

SUBJECT:

Second supp) to 820407 application for Cycle 4 reload 8 uprate license amend, Further changes to App A Tech Specs necessary to allow Cycle 4 operation w/Exxon Nuclear Co fuel at 3i411 MHte DI'STRIBUTION CODE: A001S TITLE: OR COPIES RECEIVED:LTR Submittal: General Distr ibution g ENCL J SIZE:, Jg NOTES<<

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL NRR ORB1 BC, 01 7 7 INTERVAL; ELD/HDS3 1 0 NRR/DHFS DEPY08 1 1 NRR/DL DIR 1 1 N ORAB 1 0 NRR/DS I/RA8 1 1 G F IL 04 1 1 RGN3 1 1 EXTERNAL: ACRS 09 6 6 LPDR 03 1 NRC PDR 02 1 1 NSIC 05 1 1 NTIS 1 1 TOTAL NUMBER OF COPIES REQUIRED: LTTR 24 ENCL 22

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INDIANA 5 MICHIGAN ELECTRIC COMPANY P. O. BOX 18 BOWI.ING GREEN STATION NEW YORK, N. Y. 10004 September 30, 1982 AEP:NRC:0637E Donald C. Cook Nuclear Plant, Unit No. 2 Docket No. 50-316

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License No. DPR-74 UNIT 2 CYCLE 4 TECHNICAL SPECIFICATIONS CHANGE REQUEST Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation

. U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Dear Mr. Denton.

This letter and its Attachments constitute the second supplement to the "Application for Unit 2, Cycle 4 Reload and Uprate License Amendment" contained in our letter No. AEP:NRC:0637A dated April 7, 1982. The first supplement to our application is contained in letter No. AEP:NRC:0637B, dated July 8, 1982.

This letter requests further changes to the Unit 2, Appendix A Technical Specifications (T/S) . These changes are necessary to allow Cycle 4 operation with Exxon Nuclear Company (ENC) fuel at the slightly higher reactor thermal po~er of 3411 MWt.

'I Two Attachments are included in this letter in support of our application. Attachment No. 1 contains a description of the proposed T/S changes. Attachment No. 2 contains the revised T/S pages, All changes in Attachment No. 2 are indicated by a vertical line on the right<<hand side of the page. The T/S changes contained in Attachment No. 2 to this letter have been reviewed and approved by the Plant Nuclear Safety Review Committee (PNSRC) and by the AEPSC Nuclear Safety and Design Review Committee (NSDRC).

As a result of an issue raised by another fuel supplier in 1981, an additional LOCA analysis is being performed with ENC's EXEM/PWR evaluation model to calculate peak cladding temperature with maximum Emergency Core Cooling System (ECCS) safeguards in operation. As indicated in our letter No. AEP:NRC:0637D, dated September 8, 1982, the results of this additional analysis are scheduled to be submitted to the NRC Staff by November 1, 1982, ENC has informed us that certain T/S values will change as a result of this additional analysis. In the interest of avoiding a duplication of effort, those T/S changes which are dependent on the ECCS reanalysis results are not presented here.

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Lg Mr. Harold R. Denton AEP: :0637E Upon completion of the ECCS reanalysis, the final version of the affected T/S will be submitted.

Payment for NRC processing of the aforementioned requests has previously accompanied our "Application for Cycle 4 Reload and Uprate License Amendment" contained in letter No. AEP:NRC:0637A wherein we interpreted this application for a license amendment to constitute a Class IV Amendment as defined in 10 CFR 170.22.

4 The T/S changes requested in Attachments No. 3, 4 and 6 to our letter No. AEP:NRC:0637A, the requests contained in this letter, and the requests to be submitted once the ECCS reanalysis described above is completed, constitute all the T/S changes necessary for the operation of Unit 2 during Cycle 4. Please note that the T/S change requests contained in Attachments No. 3 and 4 to AEP:NRC:0637A are still outstanding.

Our review of the Unit 2, Appendix B T/S has concluded that no changes need to be submitted for them.

In our letter No. AEP:NRC:0637B dated July 8, 1982, we informed you that ENC had documented the results of their ECCS analysis as it relates to ENC fuel, and of the more limiting transients for the Cook Plant, in Reports No. XN-NF-82-35 ("Donald C. Cook Unit 2 LOCA ECCS Analysis Using EXEM/PWR Large Break Results" ) and No. XN-NF-82-32

("Plant Transient Analysis for the Donald C. Cook Unit 2 Reactor at 3425 MWt"), respectively. ENC has transmitted to the NRC twenty-five (25) copies of each of these reports along with twenty-five (25) copies of Report No. XN-NF-82-37, entitled "D. C. Cook Unit 2, Cycle 4 Safety Analysis Report" under separate cover. In addition, ENC has recently issued Supplement 1 to Report No. XN-NF-82-37 and has transmitted twenty-five (25) copies of this supplement under separate cover to the NRC. We request that you incorporate all of the above-mentioned reports onto Docket No. 50-316 for Unit No. 2 of the Donald C. Cook Nuclear Plant.

This document has been prepared following Corporate Procedures which incorporate a reasonable set of controls to insure its accuracy and completeness prior to signature by the undersigned.

Very truly yours, R. S. Hunter

/os Vice President cc:(attached)

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Mr. Harold R. Denton AEP: :0637E cc: John E. Dolan - Columbus M. P. Alexich R. W. Jurgensen W. G. Smith, Jr. Bridgman R. C. Callen G. Charnoff Joe Williams, Jr.

NRC Resident Inspector at Cook Plant - Bridgman

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ATTACHMENT NO. 1 TO AEP:NRC:0637E DONALD C. COOK NUCLEAR PLANT, UNIT NO. 2 DESCRIPTION OF UNIT 2 CYCLE 4 TECHNICAL .SPECIFICATION CHANGE REQUESTS

Chan e No. 1 Pages 2-2 and 2-3; Figures 2.1-1 and 2.1-2 Core safety limit lines for the D. C. Cook Unit 2 reactor have been regenerated to include tPe effects of Exxon Nuclear Company's (ENC)

ZNB CHF correlation and a F'j limit of 1.60. The safety limit lines define reg'ions of reactor operation in terms of the system pressure,

,vessel avezage coolant temperature, and core power level. Operation within the safety 1Mit lines ensures 'that the vessel hot leg temperature will be less than the saturation temperature and that the min&urn departure from nucleate boiling ratio (MDNBR) will be greater than the safety limit value of 1.17 as calculated with the ZNB critical heat flux correlation. These core safety limit lines are shown in Figure 2.1<<1 for, four-loop operation and in Figure 2.1-2 for three-loop operation.

-- It should be noted that the Core Safety Limit lines presented herein reflect the change in calculated DNB margin resulting from use of a measuzed non-mixing vane spacer loss coefficient in the ZNB critical heat flux calculation. Previous ENC-generated Core Safety Limit lines, such as those presented in INC Report No. ZN-NF-82-32(P) entitled "Plant Transient Analysis for the Donald C. Cook Unit 2 Reactor at 3425 MPt",

employed a loss coefficient calculated from the generalized de Stordeur equation.

Chan e No.. 2 Page 2<<8; Table 2.2-1 The analyzed increase in F N and the relative thermal margin detXodt for ENC fuel require ther the Pverteepereture QT trip eetpoiut be reduced. The trip setpoint equation is identical to that currently given in the T/S. However, certain parameters which appear in this equation are required to be changed. The bias constant K1 in the trip setpoint equation for 4-loop operation has been reduced in magnitude by the equivalent of 4% in power. The pressure gain K3 for both 3-loop and 4-loop operation has also been ad)usted to better correlate the

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necessary pressuze dependence of the trip setpoint. The change to the lMitation on indicated T at Rated Thermal Power has previously been submitted.

The Overtemperature 4T trip setpoint employed in the Plant Transient Safety (PTS) analysis conservatively represents the recommended changes. The validity of the recommended setpoint equation for 4-loop operation has been further tested with steady state thermal margin calculations over the temperature, pressure, and power range of interest. Both the PTS analysis reported in ZN-NF-82-32(P), "Plant

Transient Analysis for the Donald C. Cook Unit 2 Reactor at 3425 MWt",

and the steady state MDNBR calculations, fully support the inclusion of the recommended 4-loop Overtemperature DT trip equation in the reactor protection system.

The gain constant in the Overtemperature +T trip equation for 3-loop operation is the same as that, in the 4-loop trip equation. The current T/S bias constant K> for 3-loop operation has been retained.

The adequacy of the 3-loop Overtemperature Zh,T trip equation, incorporating the new gain constant, has been tested with steady state thermal margin calculations over the temperature, pressure, and power ranges of interest. The results of these test calculations confirm the conservatism of the Overtemperature DT trip equation for 3-loop.

operation.

Chan e No. 3 Page B 2-1; Specification 2.1.1 The basis 'for reactor Core Safety Limits has been 'zevised to reflect ENC's thermal margin methodology. The use of the XNB critical heat flux correlation and its associated safety limit is specified. The fourth paragraph . has been deleted, since ZNC has not employed a statistical combination of uncertainties in its thermal margin analyses for D, C. Cook Unit 2. Appropriate uncertainties in the thermal margin analysis have been treated deterministically.

ATTAQQKNT NO ~ 2 TO AEP:NRC:0637E DONALD C. COOK NUCLEAR PLANT, UNIT NO. 2

-REVISED TECHNICAL SPECIFICATION PAGES

660 UNACCEPTABLE PERATION 640.

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.4 .6 .8 1.2 FRACTION OF RATED THERMAL POWER Figure 2.1-1 Reactor Core Safety Limits-Four Loops in Operation D. C. COOK UNIT 2 2-2 AiiENDNEVZ M3 ~

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ACCEPTABLE CL O OPERATION 560 0 .? 1.0 FRACTION OF RATED THERMAL POWER Figure 2. 1-2 Reactor Core Safety Limit-Three Loops in Operation D. C. COOK - UNIT 2 2-3 AHEÃ)MENT NO.

Cl nC) TABLE 2.2-1 (Continued)

C) 7C REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION Continued)

Operation with 4 Loops Operation with 3 Loops 1.116 K1

= 0.01607 = 0.01607 K2 K2

= 0.000926 = 0.000926 K3 K3 I

CO and f, (hl) is a function of the indicated difference between top and bottom detectors of thd power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(i) for - between - 40 percent and + 3 percent, q q f> (al) = 0 (wherk q knd q are percent RATED THERI(AL POHER in the top and bottom halves of the c re respectively, and qt +

qb is total THERHAL POHER in percent of RATED THERI1AL POWER).

(ii) for each percent that the magnitude of (q - q ) exceeds - 40 percent, the hT trip setpoint shall be automatical)y reIIuced by 1.8 percent of m its value at RATED THERMAL POWER.

C3 (iii) for each percent that the magnitude of (q - q ) exceeds + 3 percent, the hT trip setpoint shall be automatical)y reIIuced by 2.2 percent of C) its value at RATED THERMAL POWER.

2.1 'AFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant satura-tion temperatures Operation above the upper boundary of the nucleate boi ling regime could result in excessive cladding temperatures because of the onset of

, departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temper-ature and Pressure have been related to DNB. This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat

'flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

, The DNB design basis is as follows: there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during, Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the XNB correlation in this applica-tion). The correlation DNBR limit is established based on, the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit.

The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature below which the calculated DNBR is no less than the correlation DNBR limit value or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid. Uncertainties in primary system pressure, core temperature, core thermal power, primary coolant flow rate, and fuel fabrication tolerances have been included in the analyses from which Figures 2.1-1 and 2.1-2 are derived.

D. C. COOK - UNIT 2 ,B 2-1 'MENDMENT NO.