ML20248A532
| ML20248A532 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 07/21/1989 |
| From: | Sieber J DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20248A538 | List: |
| References | |
| NUDOCS 8908080307 | |
| Download: ML20248A532 (3) | |
Text
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$happmgport. PA 4G77 0004 Vem Preso n Nuclear Gung, '
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July 21, 1989 eha
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H U.--S. Nuclear Regulatory Commission li
-Attn:
Document Control-Desk L
Washington, DC_ 20555
Reference:
Beaver Valley Power Station', Unit No. 1 Docket No. 50-334, License No. DPR-66 Technical Specification Change No. 162 Additional Information Gentlemen:
Attachment A
provides our response. to a
request for information' from an NRC staff member during a
telephone
' conversation _ dated July 17, 1989.
This information is provided in support of Technical Specification -Change' No.
162 concerning reanalyses of-UFSAR Section 14.2.7 " Single Reactor Coolant Pump Locked Rotor".
The staff requested that we revise the design bases.of UFSAR Section 14.2.7 to evaluate those fuel rods reaching DNB and_the resulting radiological consequences.
The radiological evaluation has been completed-for the locked rotor accident assuming those rods that reach DNB-fail and is 'provided in I
g Attachment B for your review.
Should you have additional questions, please contact my staff.
L Very truly yours, JLh4--
D. Sieber ice President Nuclear Group cc:
.Mr. J.
Beall, Sr. Resident Inspector Mr. W. T. Russell, NRC Region I Administrator Mr.
P. Tam, Sr. Project Manager CV Ult
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ATTACHMENT A D
- e 1...The Lstatisticaliconvolution method described in WCAP-10125 for-Lthe evaluation of initial fuel rod.to nozzle growth gap has M
not been : approved.
This method should not be.used- 'in VANTAGE 5.
Response:'
Werst~ case fabrication tolerances and fuel rod and assembly g; eth.are used to ' determine the initir.1: fuel rod to nozzle gru 2 gaps-in the evaluation of fuel rod. performance sumurized in.Section 2.4 of the " Plant Safety Evaluation for Beaver _ Valley: Power Station _ Unit 1 Fuel Upgrade and Increased-Peaking Factors."
This.is in compliance with condition 1 of the VANTAGE 5 NRC Safety Evaluation Report.
1<
H 2.
'If a
positive' Moderator Temperature-Coefficient (MTC) is intended' for VANTAGE 15, the same positive MTC consistent with L
,the plant' Technical Specifications should be used in the plant specific safety analysis, i
Response::
1' The. current licensing basis and the proposed operating license change request No. 162 submittal do not use a positive MTC.
3.
Address additional FSAR changes for the Locked Rotor accident.
Response
FSAR- ' pat F' changes for Section 14.2.7 Single Reactor Coolant Pump Locked Rotor were included-in the
" Plant Safety Evaluation for Beaver Valley Power Station Unit 1 VANTAGE SH Fuel Upgrade and Increased Peaking Factors."
Additional UFSAR. changes will be incorporated in the next UFSAR update to address the change in basis and radiological consequences for the locked rotor accident.
4.
What was the approved method for determining the percent of fuel urods in DNB for the Locked Rotor accident and what are the results?
Response
An' analysis was performed to determine the percent of fuel rods in DNB for the Locked Rotor accide.nt.
The coolant conditions were calculated with the THINC-IV computer code as specified in the Beaver Valley Power Station 1 PSAR.
The MINI-RTDP DNB methodology described in
" Plant Safety Evaluation for Beaver Valley Power Station Unit 1 VANTAGE SH Fuel Upgrade and Increased Peaking Factors" was used to evaluate the DNB acceptance criterion.
1 I
t ATTACHMENT A Page 2 j
It was found that 18% of the fuel ?ods would be in DNB with
)
minimum DNBRs-less than the safety aualysis PNBR limit.
This i
. calculation is based on a
fuel rd power census which is j
conservative for Cycle 8
ortratior and is expected to bound 1
all future cycles.
The radiological consequences of a single reactor coolant pump locked rotor were analyzed to show compliance with the siting guidelines of 10 CFR 100 using methodology consistent with l
Standard Review Plan (SRP) 15.3.3 - 15.3.4.
The percent of l
fuel rods projected to experience DNB was determined to be 12.5%
based on the cycle 8 fuel rod power census and 18% is i
expected to conservatively beund future fuel cycles.
Although the peak clad temperature of the fuel will not exceed 2700*F, all of the fuel rods experiencing DNB are conservatively I
assumed to fail.
The radiological analysis assumes the instantaneous release of 18% of the gap activity to the RCS.
A coincident loss of AC power to station auxiliaries is
- assumed, resulting in an 8-hour plant cooldown via steam release from the secondary system to.the atmosphere.
Technical specification primary-to-secondary leakage and initial RCS and steam generator activities are assumed.
The analysis projected 0-2 hour doses at the exclusion area i
boundary of 2.3 rem whole body, 1.4 rem beta skin, and 21.6 i
rem thyroid.
The dose results of the analyses are a small l
fraction (i.e.,
less than 10%)
of the 10 CFR 100 exposure guidelines and
- are, therefore, acceptable.
The projected
)
doses are within 10 CFR 100 siting exposure guidelines even if the accident should occur coincident with an iodine spike.
l The analysis projected control room doses for the duration of i
the accident of 0.22 rem whole body, 2.9 rem beta skin, and 19.1 rem thyroid.
The dose results are within the criteria of j
GDC-19 and are, therefore, acceptable.
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