ML20238E923
| ML20238E923 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/09/1987 |
| From: | Mroczka E NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20238E926 | List: |
| References | |
| B12661, NUDOCS 8709150294 | |
| Download: ML20238E923 (7) | |
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- NORTHEAST UTILITIEE ceners Omces. seoen street. Bemn, connecticut EdNNNbEtNe P.O. BOX 270 xes wrooma co""*
H ARTFORD, CONNECTICUT 06141-0270 J 7.,$EN,,*"ffy",
(203) 665-5000 k
k September 9,1987 Docket No. 50-423 B12661 Re: 10CFR50.90, U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555
Reference:
(1) Reload Safety Evaluation for Millstone Unit No.
3, Westinghouse Electric Corporation, August,1987.
(2) Positive Moderator Temperature Coefficient Licensing Report for Millstone Unit No. 3, Westinghouse Electric -
Corporation, August,1987.
(3) RTD Bypass Elimination Licensing Report for Millstone Unit No.
3, WCAP-Il496 (Proprietary),. WCAP-ll497 (Non-Proprietary), June 1987.
Gentlemern Millstone Nuclear Power Station Unit No. 3 Cycle 2 Reload, Technical Specification Change Request Pursuant to 10CFR50.90, Northeast Nuclear Energy Company (NNECO) hereby proposes to amend its Operating License NPF-49 by incorporating the attached proposed changes into the Technical Specifications of Millstone Unit No. 3.
A description of the proposed changes is provided in Attachment 1.
The revised pages of the Technical Specifications are provided in Attachment 2.
The Reload Safety Evaluation Report (Reference (1)) presents an evaluation for Millstone Unit No. 3, Cycle 2, which demonstrates that the core reload will not adversely affect the safety of the plant for either four-or-three-loop operation.
This evaluation was accomplished utilizing the methodology described in WCAP-9273-A, " Westinghouse Reload Safety Evaluation Methodology." Included in this l
evaluation is consideration of Resistance Temperature Detector (RTD) Bypass Elimination (Reference (3)) and operation with a positive moderator temperature coefficient (PMTC) (Reference (2)) at reactor power levels less than 100E As requested by the NRC Project Manager for Millstone Unit No. 3, Mr. R. L.
Ferguson, four_(4) copies of the technical reports, References (1) and (2), are being forwarded directly to him.
Since Reference (3) contains information proprietary to Westinghouse Electric Corporation, it will be submitted under a separate cover.
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B12661/Page 2 September 9,1987,
- NNECO. has reviewed the proposed changes pursuant to 10CFR50.59 and has -
2 determined'that they constitute an Unreviewed Safety question due to a small increase in the radiological consequences of a Locked Rotor Accident for three-loop operation, but has determined the proposed changes to be acceptable and safe. The basis of the determination is discussed in Attachment 3.
NNECO has also reviewed the proposed changes in accordance with 10CFR50.92 arid has concluded that they do not involve a significant hazards consideration because the changes do not:
1.
Involve a significant increase in the probability or consequences of an accident previously evaluated.
As stated in ' Attachment 3, the only accident.for which the radiological consequences are affected is by the proposed changes to the allowable moderator temperature coefficient for the Locked l Rotor Accident for three-loop operation.
The calculated increases in radiological consequences are not significant.
Table I to provides the radiological consequences due to the Locked
. Rotor Accident for three-loop operation. The 0-2 HR EAB thyroid dose only increases by 1 Rem or 0.3% of the 10CFR100 limit. The 0-2 HR EAB whole body dose only increases by 0.32 Rem or 2% of the 10CFR100 limit.
The LPZ thyroid dose only increases 3.1 Rem or 1% of the 10CFR100 limit.
The LPZ whole body dose increases by 0.047 Rem or 0.2% of the 10CFRI00 -
limit. Since the above increases in consequences are not significant, the proposed changes do not involve a significant hazards consideration. In addition, the proposed changes will not have any impact on the probability of occurrence of any' design basis accident.
2.
Create. the possibility' of a.new or different kind of accident from any previously evaluated. As stated in Attachment 3, there are no new failure rnodes associated with these changes and the impacts are all covered by the existing design bases. As discussed in Reference 1, it is possible that plant response to transients will change due to the positive moderator temperature coefficient (PMTC). Some changes to control systems may be required althgugh it is not expected.. This will not affect the safety analysis or create a new accident since credit for control systems is not assumed in the PMTC re-analyses. The proposed changes, resulting from the replacement of the RTD bypass system with thermowell-mounted RTDs, will not physically affect the performance or reliability of any protection or control system. There are no new failure modes associated with these changes.
l 3.
Involve a significant reduction in a margin of safety.
As stated in l, the consequences of transients reanalyzed for the PMTC l
have not degraded to the point of reducing the margin of safety. Also, L
from a chemistry, corrosion and material compatibility standpoint there were no concerns identified.
Therefore, there is no effect on the protective boundaries. In addition, the proposed changes, resulting from the elimination of the RTD bypass system, have no adverse impact on the basis of any Technical Specification. The only changes are to the specific values of the parameters contained in the Technical Specification.
Therefore, the proposed changes do not reduce the margin of the safety as specified in the basis of any Technical Specification.
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N i Ui S2 Nuclear Regulatory Commission -
- B12661/Page 3 September 9,1987
'Moreover,L the Commission has provided guidance concerning the application of
- standards in 10CFR50.92 by providing certain examples (March 6,1986, FR,7751) of amendments that at considered not likely to involve a significant hazards consideration. The ctR 4 proposed herein are enveloped by example.(vi),.a
. change that either results in some increase to the probability,or consequences of a previously analyzed.sccident or may reduce in some way a safety ;nargin, but where the results of the change are clearly within all acceptable criteria with respect to systems or components specified in the Standard Review: Plan.. As
- stated above, the only accident for which the radiological consequences-are affected is by the proposed changes to the allowable moderator temperature.
coefficient for the Locked. Rotor Accident for three-loop operation. Since, the increase in consequences '(Table 1 to Attachment.3) is not significant, the proposed changes do not involve a significant hazards consideration.
Based upon the information contained in this submittal and the environmental assessment 'for~ Millstone' Unit.No. 3, there are no significant radiological or nonradiologicalimpacts associated with the proposed action and that the proposed license amendment will not have a significant effect on the quality of the human environment.
.The' Millstone' Unit No. 3 Nuclear Review Board has reviewed and approved the attached proposed revision and concurs with the above determinations.
- In accordance with 10CFR50.91(b), NNECO is providing the State of Connecticut with a copy of this proposed amendment.-
Pursuant to the requirements of 10CFR170.12(c), enclosed with this amendment l
request is the application fee of $150.
We trust you find this information satisfactory and request review and approval of this amendment request by Decemeber 5,1987, in order to support the start of Cycle 2.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY E.7 roczka (/
Se r Vice President cc Mr. Kevin McCarthy l-Director, Radiation Control Unit l
Department of Environmental Protection L
Hartford, Connecticut 06116 W. T. Russell, Region I Administrator R. L. Ferguson, NRC Project Manager, Millstone Unit No. 3 W. J. Raymond, Resident inspector, Millstone Unit No. 3 L_=_-_____________:_--_____=__-
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B12661/Page 4 o
' ' September 9,-19875
' STATE OF CONNECTICUT)
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_l Then personally' appeared before me, E. 3. Mroczka, who being duty. sworn, did state that he is Senior Vice President of Northeast.' Nuclear Energy Company, a -
Licensee herein,.that' he is. authorized to' execute ' and-file _the foregoing -
information in the 'name and on behalf of.the Licensees herein, and that the -
Statements contained in said information are true and correct to the best of his j
- knowledge and belief.
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B12661 Attachment I c
l Millstone Nuclear Power Station, Unit No. 3 Description of Proposed Technical Specification Changes i
Cycle 2 l'
September,1987
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. Attachment 1 Description of Proposed Technical Specification Changes The proposed Technical Specification changes have been prepared to support the Cycle 2.. reload. These changes will allow a positive. moderator temperature coefficient (PMTC) at reactor power levels less than 100% In addition, the changes proposed herein are required as a result of the project to replace the existing resistance temperature detector (RTD) bypass system with.thermowell-mounted RTDs. The proposed changes are discussed below.
A.
Technical Specification Changes due to Cycle 2 Reload 1.
Section 3.1.1.3 - The proposed change would allow a +5 pcm/0F MTC below 70 percent of rated power, ramping down to O pcm/0F at 100 percent power.
2.
Section 3.1.2.5, 3.1.2.6 and 3.5.4_ - The proposed change will increase the range of acceptable boron concentration in the Refueling Water Storage Tank (RWST) to 2300 - 2600 ppm from the previous range of 2000-2200 ppm.
3.
Section 3.5.1 - The proposed change will increase the range of acceptable boron concentration in the Accumulators to 2200 - 2600 ppm from the previous range of 1900 - 2200 ppm.
te.
Section 3.6.2.3 - The proposed change will increase the range of acceptable sodiurn hydroxide concentrations in the chemical addition tank (CAT) to 2.41 - 3.10 % from the previous range of 1.35 - 2.00 %
The volurne (level) in the CAT is reduced to a range of 18000 - 19000 gallons from the previous range of 19100 - 20100 gallons.
3.
Section 3.9.1
- The proposed change will increase the boron concentration in the filled portion of the reactor coolant system (RCS) and refueling canal during Mode 6 to correspond with the new minimum RWST boron concentration.
B.
Technical Specification Changes due to the Elimination of the RTD Bypass System 1.
Table 2.2 The proposed change will revise the values of 'Z' and Sensor Error (s) for the overtemperature AT and overpower AT trips.
In addition, the proposed change will revise the values of 'Z', the sensor error (s) and the allowable value for the reactor coolant flow-low trip. In the Table notations for Table 2.2-1 changes are proposed l
to note #1 to specify how AT is to be measured, and to change the l
lead-lag compensator time constant 41. Notes #2 and #4 are to be revised to change the amount by which a channel maximum trip setpoint may exceed its computed trip setpoint.
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L Attachment l' B12661/Page 2 l 2.
Sections 3.2.3.1-and 3.2.3.2 - The proposed change -will revise-the minimum RCS flow' rates for both four and three-loop operations. -
The uncertainty values for flow measurement specified in Sections
- 1 3.2.3.1 and 3.2.3.2 are also proposed to be changed.
3.-
Table 3.2 The proposed change wlll revise the values specified for.
lj RCS Tave an6 Pressurizer Pressure which are the limits assumed.in the safety analysis.
4..
Table 3.3 2 - The proposed change will revise the response time for the overtemperature AT and overpower AT trip functions.
' 3.
' Table 4.3-1 The. proposed change ' will. delete the reference to footnote #12 in Table 4.3-1, since it is no longer relevant once 'the -
RTD bypass system is removed.
6.
Table 3.3 The proposed change will revise the allowable values for.
functional unit 5.d.1 and 5.d.2, "Tave low coincident with reactor trip (P-4)" and Func'tional Unit 9.b, "ESFAS Interlocks'- Low Low Tave (P-12)."
7.
Table 3.3 The proposed change will revise the response time for
-feedwater isolation on Tave low coincident with reactor trip (P-4).
C.
Other Technical Specification Changes The_ following Technical Specification changes required for Cycle 2 start-up will be submitted to the NRC on or about September 30,1987.
1.
S_ection 5.3.1 - The proposed change will delete the maximum total weight of uranium, limitation per fuel rod. Also, the proposed change clarifies the maximum enrichment for future core loads.
2.
Section 3.9.I'- Adds a new section 3.9.1.2 which provides an 800 ppm boron-minimum concentration in the spent fuel pool whenever fuel movement is in progress.
3.
Section 3.h.l.6 - In some areas the existing specification would require that an isolated loop be greater than 2300 ppm prior. to bringing it back into service. The new specification will req'uire at most 2300 ppm.
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