View source for W3P86-3384, Forwards Response to NRC Request for Addl Info Re Part B of Cycle 2 Reload Analysis,Submitted on 861001

{{Adams
| number = ML20215M801
| issue date = 10/28/1986
| title = Forwards Response to NRC Request for Addl Info Re Part B of Cycle 2 Reload Analysis,Submitted on 861001
| author name = Cook K
| author affiliation = LOUISIANA POWER &amp; LIGHT CO.
| addressee name = Knighton G
| addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| docket = 05000382
| license number = 
| contact person = 
| document report number = W3P86-3384, NUDOCS 8611030306
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC
| page count = 4
}}

=Text=
{{#Wiki_filter:e o
B
      $IE0NiYSW l OUISIANA POWER    & LIGHT/ 317 NEW BARONNESTREET
                                                            +  P. O. BOX 60340 ORLEANS, LOUIS!ANA 70160   +   (504) 595-3100 October 28, 1986 W3P86-3384 A4.05 QA Mr. George W. Knighton, Director PWR Project Directorate No. 7 Division of PWR Licensing-B Office of Nuclear Reactor Regulation Washington, D.C. 20555

==SUBJECT:==
Waterford SES Unit 3 Docket No. 50-382 Reload Analysis Report (Part B)
Additional Information

==REFERENCE:==
W3P86-3328 dated October 1, 1986

==Dear Mr. Knighton:==

By the referenced letter LP&L submitted Part B of the Cycle 2 Reload Analysis Report (RAR). This part of the RAR contained new sections 7.0, 8.0 and 10.0 as well as supplements to sections 4.0 and 12.0.
In subsequent discussions, your staff has requested additional information concerning this submittal. Enclosed please find our response.
!     Should you require any further information please contact Mike Meisner at I
(504) 595-2832.
Yours very truly, edad K.W. Cook Nuclear Support & Licensing Manager KWC/MJM/DPS/pim cc:   B.W. Churchill, W.M. Stevenson, R.D. Martin, J.H. Wilson, L. Kopp (NRC/NRR),
j           NRC Resident Inspector's Office (W3) 8611030306 861028 PDR P
ADOCK 05000382 PDR                                                O "AN EQUAL OPPORTUNITY EMPLOYER"                                     l

Att:chment to W3P86-3384
,                                                                        Page 1 of 3 Responses to Requests for Additional Information Reload Analysis Report Question:
Why is the limiting inside containment steam line break size only 5.0 sq. ft.
for the Cycle 2 analysis whereas 7.88 sq. ft, was assumed in the Reference Cycle analysis?

===Response===
The evaluation of pre-trip fuel failure performed for the Reference Cycle is reported in Appendix 15.C.2 of the WSES-3 FSAR. It assumes that offsite power is lost concurrent with the initiation of the accident. These differences in the postulated accident conditions cause the difference in the limiting break size.
The smaller break size delays the time of reactor trip which results in a greater potential for fuel failure.
2 The identification of 5.0 ft as the limiting inside containment break for the calculation of pre-trip fuel failure was the result of a parametric analysis in break area, time-in-cycle, and time of loss of offsite power. This analysis follows the guidance of SRP 15.1.5.
Question:
Was new ECCS-large break evaluation model, which was approved on July 31, 1986, by letter from D. Crutchfield to A. Scherer, used in the Cycle 2 LOCA analysis?
If so, was a spectrum of break sizes evaluated? Also, confirm that the assump-tion of no single failure involving the coolant injection pumps is the worst case (results in the highest peak cladding temperature).

===Response===
The Cycle 2 LOCA analysis uses the new ECCS methods which the NRC approved July 31, 1986 except that the blowdown hydraulics calculation (CEFLASH-4A) was not repeated. Section 8.2 of the Cycle 2 Reload Analysis Report describes the large break LOCA methodology utilized. The methodology was previously identified to the NRC via letter W3P86-1578 from K.W. Cook (LP&L) to G.W. Knighton (NRC), dated July 1, 1981. Except for the assumption of no single failuro and a different axial power shape (consistent with the methodology approved by the NRC on July 31, 1986), the Cycle 2 methodology is the same as that used for Cycle 1. Therefore, the Cycle 2 evaluation is limited to the reanalysis of the limiting break size determined by the Cycle 1 break sensitivity study and utilizes the CEFLASH-4A calculation of Cycle 1.
An assumption of no single failure (i.e., no LPSI pump failures) yields a higher peak cladding temperature because of the lower containment back pressure caused by increased spillage from the safety injection system. Lower containment and system pressures increase steam binding effects, reducing core reflood rates and core heat transfer. Cycle 1 analyses indicated that the reactor vessel downcomer NS41215

Attachm nt to
  .                                                                         W3P86-3384
.                                                                           Page 2 of 3 Response (Cont'd.):
annulus filled before the safety injection tanks emptied, and remained filled with only one low pressure safety injection pump (LPSI).       Therefore, the flow from a second LPSI pump does not benefit the core cooling. It increases the amount of spillage and contributes to lowering the containment pressure.
Question:
Does the analysis of the total loss of forced reactor coolant flow assume the trip occurs then the reactor coolant flow reaches 96.5% of its initial value or when the coolant pump shaft speed reaches 96.5% of its initial speed?

===Response===
As described in the sequence of events table, 7.3.2-2, the analysis of the total loss of forced reactor coolant flow assumed that the trip occurs when the reactor coolant pump shaft speed reaches 96.5% of design speed.      This sequence of events is consistent with the CPC algorithms and data base which will be implemented for Cycle 2.
Question:
The peak linear heat generation rate for the CEA withdrawal event from subcriti-cal conditions was calculated to be greater than the steady state centerline melt limit of 21 kw/ft. Describe the method used to calculate the fuel centerline temperature and the value calculated.

===Response===
The energy produced during the CEA withdrawal from subcritical. conditions was found by graphically integrating the resultant power excursion. The total energy production was 4850 MW-sec, which is about 1.4 full power seconds.
The maximum centerline enthalpy of the fuel is defined as the initial enthalpy plus that added during the transient. The calculation assumed that no heat is transferred away from the centerline during the transient (i.e. , adiabatic conditions).
The deposited enthalpy was calculated as follows:
EDEPOSITED = Core Power x Time at Power x 3-D Peak Mass of UO2 Present Substituting and using appropriate conversion factors:
3 E
DEPOSITED  = 4850 MW-sec x 7 x 239 Cal x 10 Kw 94.3 x 10''gm     kw-sec      MW
                 = 86 cal /gm NS41215

Attach:.nnt to
  .-                                                                       W3P86-3384
.                                                                          Page 3 of 3 Response (Cont'd.):
The initial enthalpy was determined based on the initial temperature. The initial temperature is 550 F which corresponds to an initial enthalpy of 16.9 cal /gm.
The total enthalpy is therefore 86. + 16.9 = 103 cal /gm. The temperature cor-responding to this enthalpy is 2590 F, which is well below the melting point of UO
* 2 Question:
The small break LOCA and the beginning of Cycle CEA ejection event initiated from zero power were_gnalyzed in the FSAR with a less positive MTC than currently requested (+0.5 x 10 delta k/k/ F). Justify why a reanalysis of these two events is not necessary for Cycle 2. Also, justify why a reanalysis of the steam generator _guberuptureisnotnecessaryinviewofthemoreneggtiveMTClimitof
     -3.3 x 10     delta k/k/ F as compared to the value of -2.5 x 10      delta k/k/ F used in the FSAR analysis.

===Response===
The ejected rod worth for the zero power case presented in the FSAR was 0.80%
delta k/k or 1.11 dollars. The comparable rod worth for Cycle 2 is 0.49% delta k/k or 0.78 dollars. This difference more than compensates for any possible additiona]4reactivityaddedbythechangeinMTCfromtheCycle1va}ueof
     +0.2 x 10 delta k/k/ F to the proposed Cycle 2 value of +0.5 x 10          delta k/k/ F for powers less than 70%. Thus the Cycle 2 results are bounded by those shown in the FSAR.
The steam generator tube rupture consequences are not sensitive to reactivity related parameters; thus, no reanalysis is required to support the more negative MTC limit.
ThesmallbreakLOCAanalysespresentedintheFSARutilizedthesamefufl power MTC value that was used for Cycle 2. For the MTC to be +0.5 x 10         delta k/k/ F, the reactor power must be less than 70% according to Tech Spec 3.1.1.3.
The small break LOCA results are more sensitive to initial power (decay heat level) than to the MTC and the short duration power change that a positive value causes before reactor trip. The reduction of the initial power more than compen-sates for a more positive MTC. In addition, the very conservative treatment of the moderator reactivity feedback characteristics assure that the small break LOCA analysis presented in the FSAR is bounding for Cycle 2.
NS41215
_}}