ML20029B307
| ML20029B307 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 02/28/1991 |
| From: | Mroczka E CONNECTICUT YANKEE ATOMIC POWER CO., NORTHEAST UTILITIES |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20029B308 | List: |
| References | |
| B13751, GL-89-19, NUDOCS 9103070031 | |
| Download: ML20029B307 (4) | |
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NORTHEAST U7?LITIES -
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February 28, 1991 Docket No. 50-213 i
B13751 Re:
10CFR50.90 GL 89-19 ISAP Topic 1.101 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Gentlemen:
Haddam Neck Plant Steam Generator Overfill Protection Proposad-Chanaes to Technical Specificatiqas.
Pursuant to 10CFR50.90, Connecticut Yankee ' Atomic Power Company (CYAPCO)
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hereby proposes to amend Operating License DPR 61 by incorporating the attached changes into the Technical Specifications for' the Haddam Neck Plant.
The revised pages are provided in Attachment No.1.
These _ proposed changes establish. periodic operability testing of the steam generator overfill protec-
-tion system at the Haddam Neck Plant.
Backaround In.'a letter -dated September 20, 1989,(I) the NRC Staff issued Generic Let-ter 89-19
" Request for Action Related to Resolution of Unresolved Safety -
Issue A-47."
As-part of the technical resolution of Unresolved Safety Issue (USI)- A-47, the NRC ' Staff concluded that all PWRs should provide automatic steam generator (SG) overfill protection and that plant procedures and techni-cal specifications for all plants. should include provisions to periodically verify the operability of the overfill-_ protection during reactor power opera-tion-(1) J.IG. Partlow letter to All Licensees, " Request for Action Related to Resolution 'of-Unresolved = Safety-Issue A-47,
' Safety Implication of Control Systems in LWR Nuclear Power Plants' Pursuant to 10 LCFR 50.54(f)--Generic. Letter 89-19," dated September 20, 1989.
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' U.S. Nuclear! Regulatory Commission-B13751/Page 2 February 28, 1991
.By letters dated March 27, 1990,(2) and July 26,1990,(3L CYAPC0 responded to Generic Letter 89-19.
As part of that response, CYAPC0 committed to submit a-license amendment request which would add pravisions to the technical specifi-cations to: periodically verify operability-- of the SG overfill protection system. The purpose of this letter is to fulfill that commitment.
Qescriotion of the Proposed Chanaes The changes proposed herein add the feedwater system isolation function to the technical specification tables for engineered safety features actuation system (ESFAS) limiting conditior,s for operation and surveillance requirements.
Specifically, the proposed caanges are:
o-Table 3.3-2, ESFAS Instrumentation (page 3/4 3-lC) -adds item 6.a. Feed-water Isolation -Steam Generator Water Level, with specifications for total number of-channels, channels to actuate, minimum channels operable,
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applicable modes,-and an appropriate action statement.
'o Table 3.3 3',
ESFAS Instrumentation Setpoints (page 3/4 3-20)- adds-Item.6.a, Feedwater isolation -Steam Generator Water Level--High, with
- specifications for trip setpoint and allowable value, o
Table 4.3-2, ESFAS Instrumentation Surveillance Requirements (page 3/4-3-22) -adds - Item 6.a, Feedwater Isolation -Steam Generator Water Level -High,-with frequencies specified for channel check, channel calibration, analog channel operational test and -trip actuating device operational-test; also specified is-the mode for which the surveillance is required.
For the SG overfill concern at the Haddam-Neck Plant, the design-basis excess feedwater event assumes-the initiating event as the feedwater regulating valve-failingLto the fully open-position. 'The' feedwater flow resulting from a fully open regulating valve is terminated by th_e operator before the affected SG is-filled. _ The operator: trips the main feedwater pumps and the reactor following L
annunciation of:the high-high SG water' level alarm from the faulted loop.
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such,-- the design -basis -analysis does not take credit for the auto _matic feed-water control system.
With the automatic-feedwater control, the -feedwater-
-regulating ' valve would normally close when wide-range SG level exceeds 69-per-cent.
(2)L E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, " Response to Generic Letter _-89-19--Request for Action Related to Unresolved Safety Issue'A-47," dated March 27, 1990.
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E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, " Supplemental Response-to-Generic - Letter 89-19--Request for Action P.el ated to e
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. Unresolved Safety Issue A-47," dated July 26, 1990.
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U.S. Nuclear Regulatory Commission B13751/Page 3 February 28, 1991 The proposed changes add the automatic feedwater isolation function for SG overfill protection into the ESFAS instrumentation tables.
By adding the feedwater isolation function to the tables (Tables 3.3 2, 3.3 3, and 4.3-2),
the feedwater isolation function will be required to be periodically, i.e., at least every six weeks, tested.
This will enhance the reliability of the SG overfill protection system.
Sianificant H gards Consideration in accordance with 10CFR50.92, CYAPC0 has reviewed the attached proposed changes and has concluded that they do not involve a significant hazards consideration.
The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised.
The proposed changes do not involve a significant hazards consideration because the changes would not:
1.
Involve a sianificant increase in the nrobability or conseauences of in accident previously evaluated.
The proposed changes consist of new technical specifications that add the periodic operability testing requirement of the steam generator overfil.1 protection system.
Adding the feedwater isolation function to the tables for ESFAS operability and surveillance requirements w!11 enhance the reliability of the overfill protection system.
No design basis accidents are affected by this change.
Therefore, there is no impact on the probability of occurrence or the consequences of any design basis events.
No safety systems are adversely affected by the change.
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Create the oo:sibility of a new or different kind of accident from any previous 1v evaluated.
Since there are no changes in the way the plant is operated, the poten-tial for an unanalyzed accident is not created.
There is no irmact on plant response to the point where it can be const ered a 'new accident, and-no new failure modes are introduced.
3, involve a sianificant reduction in marain of safety The proposed changes provide new specifications for an existing systet.
The proposed requirements do not have any adverra impact on the protr.c-L tive boundaries.
Since the proposed changes also do not affect the consequences of any accident previously analy.ed, there is no reduction l
in any margin to safety.
The Commission has provided guidance conceinirs the application of the stan-dards in 10CFR50.92 by providing certain examples (51 FR 7751, March 6,1986) of amendments that are considered not likely to involve a significant hazards consideration.
The changes proposed herein are best characterized by Exam-pie (ii), a change that constitutes an additional limitation, restriction, or l
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-U.S. Nuclear Regulatory Commission-813751/Page 4 February 28, 1991 control not presently included-in the technical specifications; e.g.,
a more stringent surveillance requirement.
As described above, the proposed changes do not-constitute-a- significant. hazards consideration -since there are. no impacts on the probability or consequences of any design basis events nor are any new -_ accidents created. -There is also no reduction' in any margin of s a fe_ty.
Based upon the =information contained in this submittal' and the environmental assessment for the Haddam Neck Plant, there are -no significant radiological-or nonradiological impacts associated with-the proposed change, and the proposed license amendment-will not have a significant effect on the quality of the human environment.
The Haddam Neck ' Plant Nuclear Review Board has reviewed and approved the-attached proposed revisions and concurs with the above determinations.
-In accordance with 10CFR50.91(b), CYAPC0 is providing the State of Connecticut with a copy _of this amendment.
There.is no specific schedule for needing the change proposed herein.-
CYAPC0 requests that this license amendment request be reviewed and issued as NRC Staff resources permit.
Very truly yours, CONNECT! CUT YANKEE ATOMIC POWER COMPANY E. J. Mp6czka
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Senior,Vice President cc: L T. Martin, Region I Administrator A. B.: Wang, -NRC Project Manager, Haddam Neck P1 ant J. T. Shediosky, Senior Resident Inspg.or, Haddam Neck Plant STATE OF CONNECTICUT)-
) ss. Berlin COUNTY OF HARTFORD ~ )
- Then-personally appeared before 'me, E. J. Mroczka, who being duly sworn, did state that he is -Senior Vice President of Connecticut Yankee Atomic-Power Company, t Licensee herein, that he is authorized to execute and file the l
foregoing information in the name and-on behalf of the Licensees herein, and l~
that the statements contained in said information are true and correct to the best of-his knowledge and belief.
b dWi NotaYy Public
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