ML20237J820
| ML20237J820 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/17/1987 |
| From: | Mroczka E, Sears C NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20237J824 | List: |
| References | |
| B12413, NUDOCS 8708260330 | |
| Download: ML20237J820 (7) | |
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(203) 665-5000 August 17,1987 Docket No. 50-245 B12413 Re: 10CFR50.90 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20535 Gentlemen:
Millstone Nuclear Power Station, Unit No.1 Proposed Revision to Technical Specifications Single Loop Operation Pursuant to 10CFR50.90, Northeast Nuclear Energy Company (NNECO) hereby i
proposes to amend its Operating License, DPR-21, by incorporating the changes identified in Attachment 1 into the Technical Specifications of Millstone Unit No. 1. provides the General Electric Single Loop Operation analysis for Millstone Unit No.1.
The purpose of these Technical Specification changes is to provide operational flexibility by allowing single recirculation loop operation at reduced power in the ever.t maintenance of a recirculation pump or other component renders one loop inoperative.
Extensive analyses have been performed to justify single loop operation.
The following briefly describes the proposed changes.
1.
One Loop Operation The current Millstone Unit No.1 Technical Specifications do not allow plant operation beyond a relatively short period (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) of time if an idle recirculatlan loop cannot be returned to service. The capability of operating at reduced power with a single recirculation loop is highly desirable from a plant availability standpoint.
2.
Operation at < 40% Power Without Forced Circulation Section 3.6.H.4 of the Technical Specifications does not permit power operation without forced circulation. This section is being modified to allow power operation (i.e., run mode) up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with core power reduced to _<_40% with no forced circulation.
3.
Determination of 3et Pump Structural Integrity During One Loop f
Operation f
Section 4.6.G.2 of the Technical Specifications requires an operator D
to perform a daily surveillance on the differential pressure of the jet gk pumps in the idle loop. The proposed Technical Specification changes i
will require an operator to perform this surveillance on the jet pumps p
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August 17,1987-in both loops. As discussed below,_this change corrects an error in the. existing Technical Specifications.
TECHNICAL BASES
' The following discusses the technical bases for these propos ed changes from three perspectives; namely, the effects on the design basis saf ety analysis, the potential for creation of an unanalyzed event, and the effects on the margin of safety.
Effect on Design Basis Safety Analysis:
1.
One Loop Operation The pressurization, loss of feedwater flow, and cecooling transients are bounded by the two loop analyses. This is because the two loop analyses are performed with core power at 100%, whereas the maximum core power allowed with single loop operation is 65%. The severity of these events is primarily dependent on the initial power level, i.e., the higher the initial power, the more severe the transient.
Since the maximum allowed power level in single loop operation is significantly lower, the two loop 100% power transients are bounding.
A calculated reduction factor of 0.86 was applied to the maximum average planar linear heat generation rate (MAPLHGR) values at full power to obtain the MAPLHGR values for single loop operation. The consequences of any size LOCA initiated from single loop operation with the reduced MAPLHGRs are bounded by the results of previously analyzed LOCAs at full power.
Additionally, the pump seizure accident is bounded by the conse-quences of a postulated LOCA. This is because both in the pump seizure and LOCA scenarios, the recirculation driving loop flow rapidly drops to zero.
However, in a LOCA, the water level decreases, while no such decrease is experienced during a pump seizure event. Therefore, the potential effects of the hypothetical pump seizure are very conservatively bounded by the effects of a LOCA and specific analyses of the pump seizure accident are not required.
The proposed Technical Specification changes will change the rod block monitor trip setting while the plant is in operation with a single recirculation pump. With the revised trip setting, the consequences of a rod withdrawal error will be no more severe than the results of the full power transient previously analyzed.
R 2.
Operation at 40% Power Without Forced Circulation Power operation without forced circulation will be allowed only for a L
limited duration ($ 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) at low power levels (140% of rated).
l The probability of a design basis transient or accident occurring during this limited duration, when the plant will be operating without forced flow, is extremely small. For example, if the plant operates without the forced flow once in a year, the probability cf an accident I.
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U.S. Nuclear Regulatory Commission B12413/Page 3 August 17,1987 during that period is less than 0.1% of the probability of the same accident occurring any time in the year.
Even if an accident were postulated during the period when the plant is operating without the forced flow, the consequences would be less limiting. By generic sensitivity studies for BWR-3s, GE has deter-mined that the initial condition of 100% core power and 100% flow provides the most limiting results for all transients. Therefore, any transient initiated when the plant is operating without forced circula-tion (with core power $_40%) will be bounded by the same transient analyzed at full power.
3.
Determination of Jet Pump Structural Integrity During one Loop Operation The current Technical Specifications (Section 4.6.G.2) require an operator to check the differential pressure of all jet pumps in the idle loop during one loop operation. The dif ferential pressure is compared to the established pattern to ensure that it does not vary by more than 10% for any of the jet pumps. If a jet pump breaks or develops a leak, its differential pressure would differ from the established pattern.
Thus, by this surveillance, the operator ensures that the recirculation flow is not bypassing the core through a broken jet pump. However, by checking differential pressure of jet pumps in only the idle loop, the operator does not determine the integrity of the jet pumps in the active loop.
These proposed Technical Specification changes correct this error by requiring an operator to perform surveillance on the differential pressure of all jet pumps. During single loop operation, with the equalizing line isolated, the idle loop experiences about 1/4 of the flow through the active loop, but in the reverse direction. Since the idle loop experiences sigmficant backflow, the operator will be able to establish a pattern of differential pressure drops for jet pumps not only in the active, but also in the idle loop.
Obviously, such an improvement in the surveillance has no adverse impact on design basis accidents.
Potential For Creation of An Unanalyzed Event:
1.
One Loop Operation The Minimum Critical Power Ratio (MCPR) safety limit is increased from 1.07 to 1.08 for single loop operation. This increase in MCPR accounts for increased uncertainty in total core flow and Traveling Incore Probe (TIP) readings associated with single loop operation.
The rod block monitor, APRM scram, and rod block trip settings are proposed to be modified for single loop operation.
The rod block monitor setting is placed to ensure that a rod withdrawal error does not result in a critical power ratio lower than the 1.08 safety limit.
The reload analyses show that a rod withdrawal error at rated power shows acceptable consequences with respect to the critical power ratio. With modified trip settings for rod blocks, a rod withdrawal error during single loop operation will not result in a critical power i
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U.S. Nuclear Regulatory Commission i
B12413/Page 4 -
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l ratio lower than 1.08, which is the safety limit. Therefore, a rod l
l withdrawal error during single loop operation does nut result in an l
I unanalyzed event. The modified APRM scram and rod block trip l
settings ensure that the plant would operate within the space defined i
by the safety analyses. In this space, GE's generic analysis shows that the 100% power and 10096 flow point yields the most limiting l
results.
The power / flow restriction for single loop operation shown in Figure 3.3-2 of the Technical Specifications ensures that the plant operates within the bounds analyzed. Also, these power and flow limits keep thermal hydraulic instability problems to a minimum.
Limiting drive flow to 5 43,000 gpm ensures that vibration limits assumed in the original design calculation are not exceeded. Limiting the core plate dP fluctuation to 2.9 psid, peak to peak, ensures consistency with the fuel channel design basis.
Another concern warranting consideration is whether the Low Pres-sure Coolant Injection (LPCI) loop selection logic would function properly if a LOCA were to occur during single loop operation.
Following a LOCA, the LPCI pumps inject flow into loop B, unless the LPCI loop selection logic senses sustained lower pressure in the preferred loop (loop B) compared to loop A. In that case, the LPCI loop selection logic assumes the break is in the preferred loop and switches the injection point to loop A. The actual " decision" that the loop selection logic makes does not occur instantly upon a LOCA signal but after a brief delay. The delay is partially determined by the operating mode (one or two loop operation) and partially by time delays in the logic system.
The logic automatically senses whether the mode of operation is normal two-loop or one-loop via differential pressure sensors measur-ing pressure across the recirculation pumps, if only one recirculation pump is running, a small break in the operable loop could potentially go undetected, ;herefore in the case of single loop operation, the operating recirculation pump is tripped and allowed to coast down before the loop selection " decision" is made. The time delay during single loop operation is achieved with a vessel pressure permissive which is set at 926 psig. This logic provides a break-size-dependent time delay which does not unnecessarily delay the large break loop selection (for breaks larger than approximately 0.5 sq. f t.) but allows adequate time for recirculation pump coast-down during small break LOCAs.
Breaks larger than approximately 0.5 sq. ft. cannot be masked by a running recirculation pump because the break flow is larger than the recirculation drive flow. Small breaks depressurize more slowly and even when the vessel depressurization is assisted by the automatic depressurization system (ADS), the time to depressurize the vessel to 926 psig is long enough to allow for substantial coast-down of a tripped recirculation pump. Therefore, it can be concluded that the operation of loop selection logic will not be degraded by single loop operation, and that all design basis accidents remain bounded by previous analyses.
U.S. Nuclear Regulatory Commission B12413/Page5 l
August 17,1987 1
l In summary, the proposed Technical Specification changes which l
would permit single loop operation continues to ensure that the plant operates within the space defined by previous analyses.
2.
Operation at < 40% Power Without Forced Circulation Allowing reactor power to be less than 40% without forced circulation does not create the potential for an unanalyzed event for the following reasons:
1 a.
Operation in this region is bounded by the 100% power /100%
flow point for all analyzed transients and accidents. There are no new, different events that can be postulated at this condition.
b.
Potential stability problems are minimized.
Past reload analyses have shown that a decay ratio at i 40% core power is f 0.6, well below the generally accepted limit of 0.8.
- Hence, no new unanalyzed accident scenarios exist.
3.
Determination of Jet Pump Structural Integrity During One Loop Operation This change only involves increasing the monitoring requirements for jet pumps during single loop operation. Changing these monitoring requirements does not create a possible unanalyzed event.
Effect on the Margin of Safety:
j l
The proposed changes have been shown not to increase the probability or i
consequences of any previously analyzed event. These changes also do not result in any unanalyzed event. The margin of safety as defined in the basis for any Technical Specification is not reduced for the following reasons:
1.
Application of a MAPLHGR reduction factor of 0.86 and increasing the I
MCPR limit by 0.01 ensures no reduction in the margins of safety of the thermal limits due to single loop operation.
2.
Modification of Rod Block and Reactor Protection System Rod Block and Scram setpoints for single loop operation ensures that the margin of safety j
for a Rod Withdrawal Error event is maintained.
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3.
Single loop power / flow restrictions were chosen to maximize attainable power levels while maintaining or increasing the margin of safety available at the most limiting two loop power / flow point (100% power /100% flow).
4.
Restricting core thermal power to less than 40% without forced circulation l
forces operation inside the analyzed Millstone Unit No. I power / flow region and thus maintains the margin of safety available at the most limiting two loop power / flow point.
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U.S. Nuclear Regulatory Commission B12413/Page 6 August 17,1987 NNECO has reviewed the attached proposed changes pursuant to 10CFR50.59 and has determined that they do not constitute an unreviewed safety question.
The probability of occurrence or the consequens 2 of an accident or malfunction of equipment important to safety (i.e., safety-related) previously evaluated in the final safety analysis report have not been increased. The possibility for an accident or malfunction of a different type than any evaluated previously in the final safety analysis report has not been created. There has not been a reduction in the margin of safety as defined in the basis for any Technical Specification.
NNECO has reviewed these proposed changes, in accordance with 10CFR50.92, and has concluded, based on the previous analyses, that they do not involve a significant hazards consideration in that these changes would not:
1.
Involve a significant increase in the probability or consequences of an accident previously evaluated.
2.
Create the possibility for a new or different kind o' accident from any previously analyzed.
3.
Involve a significant reduction in a margin of safety.
The Comrnission has provided guidance concerning the application of standards in i
10CFR50.92 by providing certain examples (51FR7751, March 6,1986). The change proposed herein most closely resembles example (vi), a change which either may result in some increase to the probability or consequences of a previously-analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan, e.g.,
a change resulting from the application of a small refinement of a previously used calculational model or design method. An editorial correction on page B 2-2 resembles example (1), a purely administrative change to the Technical Specifications.
The Millstone Unit No. I Nuclear Retiew Board has reviewed and approved the attached proposed revisions and has concurred with the above determinations.
In accordance with 10CFR50.91(b), we are providing the State of Connecticut with a copy of this proposed amendment.
Pursuant to the requirements of 10CFR170.12(c), enclosed with this amendment is the application fee of $150.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY
$.bs Act' E. J. Mroczka Senior Vice President b
N_
By: C. F. Sears Vice President
U.S. Nuclear Regulatory Commission B12413/Page 7' August 17,1987 cc:
W. T. Russell, Region I Administrator M. L. Boyle, Millstone Unit No.1 Project Manager T. Rebelowski, Millstone Unit No.1 Resident Inspector Mr. Kevin McCarthy Director, Radiation Control Unit Department of Environmental Protection Hartford, Connecticut 06116 STATE OF CONNECTICUT )
) ss. Berlin COUNTY OF HARTFORD
)
Then personally appeared before me C. F. Sears, who being duly sworn, did state that he is 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 Licensee herein and that the statements contained in said information are true and correct to the best of his knowledge and belief.
Notary Publig' Y 4?t/ /
Jhhd/E My Commission Erpires March 31,1984 i
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