ML19291B562
| ML19291B562 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/01/1978 |
| From: | Herbein J METROPOLITAN EDISON CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| GQL-0337, GQL-337, NUDOCS 7911080601 | |
| Download: ML19291B562 (13) | |
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REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
D!STRI,BUTION FOR INCOMING MATERIAL 50-289 REC: REID R W ORG: HERBEIN J G DOCD ATE: 03/01/7S NRC METROPOL EDISON DATE RCVD: 03/06/78 DOCTYPE: LETTER NOTARIZED: NO COPIES RECEIVED
SUBJECT:
LTR 1 ENCL 1 ADDL INFO CONCERNING TECH SPECS CHANGE REQUEST NO 72, CYCLE 3 EXTENSION IN RESPONSE TO NRC LTR DTD 2/24/78.
PLANT NAME:THREE MILE ISLAND - UNIT 1 REVIEWER INITI AL:
XEF DISTRIBUTOR INITI AL:
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METROPOLIFAN EDriON COMPANY w.%.-e-.-
POST OFFICE BOX 542 READING, PENNSYLVANI A 19603 TELEPHONE 215 - 929 3601 March 1, 1978 P'. Lij-GQL 0337 x
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Director of Nuclear Reactor Operations D,
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Attn:
Mr. R. W. Reid, Chief N
Operating Reacters Branch No. h
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U. S. Nuclear Regulatory Cen=ission 3
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,3 Washington, D. C.
20555
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Dear Sir:
/{,, _
Three Mile Island Nuclear Station Unit 1 {TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Thic letter is in response to your request for additional information concerning Technical Specification Change Request No. 72, Cycle 3 Extensicn. Regarding the February 2h,1978 =eeting whose representatives included =e=bers of your staff, Babcock and Wilcox, General Public Utilities and Metropolitan Edison Cenpany, enclosed are our responses to your eight questions.
Sincerely,
/
/
I
/
. G. Herbein Vice President-Generation JGE:DGM:cjg Enclosure Cf S,R 1583 240
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ADDITIONAL I'IFORMATION OtI-1 CYCLE 3 EXTENSION 1.
Why is it necessary to change the APSR position limits curve for coastdown operation? Is this change safety related or is it primarily for Cycle h burnup censiderations?
The new limits on the APSR position curve were defined using two criteria:
(1) Given an APSR position at a particular power level, the relative power peak when multiplied by the avtrage linear heat rate at that power level must give a linear heat rate less than or equal to the axially dependent LOCA limiting kv/ft given in the TMI-l Tech Spec (Figure 3 5-2J).
Ihe new limits en the APSR position curve assume that power operation up to full pcVer is possible up to 315 EFFDs.
(2) Further APSR limiting position modifications were made to assure that the limits of the other LHR-dependent parameters (core imbalance and full length rod positien) would net change from their current Tech Spec values. This criteria was achieved except for the slight change in positive imbalance presented in Figure 3.5-2I.
Also, modificatien of Figure 3.5-2N was needed in that the APSR positien for a zero imbalance point with a power coastdown vculd shift towards the +h0% vithdrawal position from the current +30%.
The original Tech Spec limit en APSR position was derived conservatively in that the limit-ing positica for the 102% power level (h9% vithdrawn) was also imposed on all pcVer levels through 605 full power. An analysis at 60% fall pcver allowed the APSR's to be ec=pletely withdrawn withcut violating LHR limits.
Plant operation never needed more APSR withdrawal between 102%
and 60% full pcver since it was always base leaded.
However, with a coast-devn, analysis was perfor=ed at several intermediate power levels, 583 241 1
vith the 1cwest level allowing full withdrawal of the APSR's.
This resulted in the visible change between Figures 3.5-2M and 3.5-2:T in the +h0% vithdrawal area.
If this detailed power dependent analysis was originally perfor=ed for Cycle 3, the change would not have been as dramatic.
As discussed above, the change was safety related (LOCA limiting kv/ft) and not to Cycle h burnup considerations.
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- 2. * '4 hat is the most reactive position for the APSR's? Is there a reactivity gain in moving them further out?
The most reactive position in the core for the APSRs at 315 EFFD's is approximately 32". vithdrawn. However, a positive imbalance would occur here.
In order to maintain zero imbalance, the APSR position would be h0-to h5". withdrawn at this core burnup. There is a slight reactivity gain in moving them further up the core, but this was not the reason for modi ^/ing Figure 3.5-2M as explained in response to the first question.
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3 We vould like a more thorough explanatien of what is being done and why.
The plant operation during the coastdown is summarized as follows.
The plant vill drop power in steps, controlling the reactivity effects of Doppler and xenon with full length control rods, while maintaining near-zero imbalance with APSR's, the ecnstant boron concentration reached' at 280 EFFDs, and a constant average moderator tenperature. The latter is attained by increasing the coolant inlet temperature as a function of the decrease in power level. Therefore, all average coolant temperature-related set points vill be naintained as prior to coastdown.
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k.
What vill be the effect en the accidents and transients as a result of repositioning the APSR's?
The results of previously limiting FSAR accident and transient analyses referenced in the TMI-1 Cycle 3 Reload Report (BA'4-lhh2) remain valid for the coastdown. A comparisen of key paraneters is given in the at-tached table. The resultant relative power peaking values during the coastdown were within the design peaking facters used in the limiting safety analyses. See Question 8 for further discussion of the affected accidents and transients.
5 1583 245
5 In BN4-3 hh2 en page h-2, the statement is made that the maximum expected three-cycle local pellet burnup is less than 55,000 ?i4DhfM.
What vill the value be for an extension of Cycle 3? Hev vill this affect the ability to meet the design criterien on cladding strain of 15?
The highest pellet burnup expected at 315 EFFD's is kh,218 Ff4D/bfM, well below the limiting 55,000 v4D/>fIU derived from the 15 clad strain criterion.
D 1583 246
6.
How does extension of Cycle 3 affect the ability to provide the required shutdown margin as shown in Table 5-2 of 3N4-1kh2?
The shutdown margin in Table 5-2 of Brd-1kh2 is derived from the vorst core condition for this parameter, i.e., prior to Group 7 rod withdrawal near the end of cycle (2h6 EFFD's). Once the control rods are removed, the maximum allovable inserted rod worth decreased, increasing the shutdown margin.
It may also be stated that after analy::ing those key pars =eters involved in this calculation that the extended 35 EFFDs of operation vill not bring the shutdown margin below the minimum value for Cycle 3 of 1.4%
Ak/k. Hence, the shutdown margin criterion of 15 Ak/k vill not be violated by extending Cycle 3's operation.
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7 How vill extension of C/cle 3 affect the minimus D:GR?
All steady state =argins to the limiting DNER are pocitive threughout Cycle 3 and, in fact, increase with cycle burnup. Therefore, the ex-tension of Cycle 3 vill not affect the existing DNER nargins.
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S.
'Inat is the effect of the Cycle 3 extensien en the accidents considered in Sections T.h, T.6, 7.8 and T.lh of BAW-lhh2?
The four accidents questioned vill be discussed separately:
a.
Cold Water (Pumo Start-uo) Accident Of the key parameters for this accident, the only one in need of discussion based en the extended Cycle 3 operatien is the moderator temperature coefficient. As seen from the attached table, the value reached is -2.6x10- A k/k F vhich is 135 less negative than the value used in the FSAR analysis. Note also that this accident is analyzed under conditions that cannot exist at TMI-1, i.e., two pump operation at 505 full power. Hence, it may be stated that the transient results vould be less severe than those reported in the FSAR.
b.
Stuck-Out, Stuck-In, or Drotted Centrol Rod The extended cycle burnup vill affect the moderator temperature coefficient, the verth of the dropped rod, and local peaking factors. The scre negative Doppler is a second order affect.
Since the moderator coefficient is less negative than that used in the analysis, the worth of the dropped rol is :ach smaller than the limiting rod vorth used, and the resultant peaking factors are vell within the design values chosen, the results of this analysis at 315 EFFD's even frca 100%
full power, if that were possible, would be =uch less severe than those frcm the analysis presented in FSAR.
c.
Steam Line Failure The key parameters to be investigated as a result of extending Cycle 3 are the moderator temperature coefficient, the worst stuck rod worth, the core's shutdevn margin, and the local power peaking factors. As discussed previously, and shown La the attached table, the FSAR analysis 1583 249 9
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used values for these parameters that bound those at 315 EFFD's in Cycle 3 Therefore, based on the above, the consequences at the steam line break in Cycle 3 would be beunded by those results presented in the FSAR.
d.
LOCA Analysis All of the related Tech Spec figures have been modified so that the LOCA linear heat rate limits vould not be violated during this extended burnup. The fissien gas pressure is also within the :TRC guidelines limits since the pin burnup reached during the cycle stretch will not exceed the analyzed burnup value wherein pin pressure equals system pressure.
1583 250 1C
COMPARISO:I CF KEY ACCIDE:!T A ID TRA:TSIEiT PARAYEEERS AFFECTED BY CYCLE 3 EXTEISICII SAFETY STUDY FOR CYCLE 3 CYCLE 3 !
PARAMETER RELOAD REPORT 315 EFFD's
-5 Doppler coeff, 10 Ak/k F (EQL)
-1,33
.152 Moderator coeff, 10- Ak/k F (EOL)
-3.0
-2.60 All rod group vorth, Tak/k, HZP 10.0 8.72 Maximum ejected rod worth, HZP, Ak/k 1.00 0.68 Maximum dropped rod worth, HFP, %Ak/k 0.h6 less than 0.20 Maximum stuck rod worth, HZP, ECC, Ak/k 2.1 2.07 Minimum shutdown margin, HZP, %Ak/k 1.0 greater than 1.h 1583 251 11