B14342, Application for Amend to License DPR-61,revising TS for RTS Instrumentation,Srs & Bases & ESFAS Instrumentation,Trip Setpoints,Srs & Bases.Draft Rev 0 to Project Assigment 90-013, CT Yankee Modernize Fwd Encl
| ML20128E673 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 01/29/1993 |
| From: | Opeka J CONNECTICUT YANKEE ATOMIC POWER CO., NORTHEAST UTILITIES |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20128E679 | List: |
| References | |
| TASK-04-02, TASK-4-2, TASK-RR B14342, NUDOCS 9302100543 | |
| Download: ML20128E673 (9) | |
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4 NORTHEAST Ui1LITIES General OHiees e Selden Street. Berlin, Connecticut 3
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k (203) 665-5000 January 29, 1993 Docket No. 50-213 B14342 Re:
10CFR50.90 ISAP Topic Nos. 1.30, 1.54, and 2.04 SEP Topic IV-2 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Gentlemen:
Haddam Neck Plant Proposed Revision to Technical Specifications Reactor Trip System Instrumentation, Surveillance Requirements and Bases; Engineered Safety Features Actuation System Instrumentation, Trio Setooints. Surveillance Reauirements and Bases Pursuant to 10CFR50.90, Connecticut Yankee Atomic Power Company (CYAPCO) hereby proposes to amend its operating license, DPR-61, by incorporating the changes identified in Attachment 1 into the Technical Specifications of the Haddam Neck Plant.
BACKGROUND By letters dated December 6, 1991,"' and March 26, 1992, t2' CYAPC0 stated that the feedwater control system upgrade will be implemented to modernize the steam generator feedwater control system.
This project will assure the appropriate isolation of sigr.als from the safety-related systems to the nonsafety-related systems for the steam flow, feedwater fl ow, and steam generator level subsystems.
This modification continues the reactor protection system (RPS) modernization effort by replacing the existing obsolete Hagan feedwater control system instrumentation aiid portions of the original RPS circuitry with a state-of-the-art upgrade.
This project will affect the following parameters and systems:
(1)
J. F. Opeka letter to U.S. Nuclear Regulatory Commission, " Integrated Safety Assessment Program," dated December 6, 1991, t
(2)
J. F. Opeka letter to U.S. Nuclear Regulatory Commission, " Isolation of Reactor Protection System From Nonsafety Systems," dated March 26, 1992.
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f U.S. Nuclear Regulatory Commission B14342/Page 2 January 29, 1993 a.
Steam Generator Feedgater Flow The existing eight nonredundant steam generator feedwater flow transmitters will be replaced by eight redundant Category 1, Class 1E transmitters (fourof the existing steam generator feodwater flow transmitters are presently used for calorimetric purposes only).
The existing loop circuitry, indicator, and recorder will also be replaced, b.
Steam Generatur Steam Flow The existing four steam generator steam flow transmitters will be replaced by eight redundant Category 1, Class 1E transmitters.
The existing loop circuitry, indicator, and recorder will also be replaced, c.
Steam Generator Pressure The existing four containment mounted steam generator pressure transmitters will be replaced by four Category 1, Class IE transmitters.
The existing loop circuitry and indicator will also be replaced.
d.
Steam Generator Narrow Ranae level The existing four steam generator narrow range level transmitters will be replaced by twelve (three channels per steam generator) redundant Category 1, Class 1E transmitters.
The existing loop circuitry, indicator, and recorder will also be replaced.
e.
Steam Line Break Flow The existing four steam line break flow transmitters will be replaced by four Category 1, Class 1E transmitters.
The existing loop circuitry and indicator will also oe replaced, f.
Reactor Trio Circuitry The reactor trip circuitry for steam flow / feed flow mismatch will be replaced with one out of two coincidence solid-state equipment.
The steam generator narrow range low level reactor trip circuitry will be replaced with two out of three coincidence solid-state equipment.
The steam line break flow reactor trip circuitry will be replaced by solid-state equipment and suitable field interface.
This upgrade will include on-line testing capability for the above reactor trip circuitry.
g.
Steam Generator Overfill Protection The existing one out of two steam generator overfill pretection circuitry from wide range level is being replaced with two out of three narrow range level circuitry.
The current overfill protection has an allowable value of s 72 percent and a trip setpoint of s 69 percent as detected by wide range level circuitry.
This will be replaced by narrow range level
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U.S. Nuclear Regulatory Commission B14342/Page 3 January 29, 1993 circuitry, with an allowable value of s 80 percent and a trip setpoint of s 74 percent.
The new overfill protection circuitry will provide automatic closure of the feedwater regulating valve, regardless of mode of operation, as well as automatic closure of the feedwater isolation motor-operated valve.
This modification affects the reactor trip system instrumentation requirements to reflect the added redundancy and the reactor trip system instrumentation surveillance requirements to reflect the on-line testing capability.
It also changes the engineered safety features actuation system instrumentation requirements to reflect narrow range steam generator level being utilized in place of wide range steam generator level for feedwater isolation.
This change depicts the increase in the total number of channels and channels to actuate, as well as the more conservative setpoint from the narrow range instrument which also employs-a two out of three coincidence feature.
This change also adds a surveillance requirement to perform a trip actuating device operational test every refueling outage.
RELATED TOPICS to this letter contains the draft final project description for the upgrade.
Much of the material provided herein is still under review.
CYAPC0 is providing this information to the Staff to support Staff review of our license amendment request.
In addition, _ as part of SEP Topic IV-2 and ISAP Topic 1.54, CYAPC0 has evaluated the impact of a mechanical failure of the main feedwater regulating valves to close on a feedwater isolation signal.
This evaluation was requested by the NRC Staff in a letter dated 0ctober 18, 1988 Since there is a need to promptly isolate the feedwater line given a high steam generator level condition, the feedwater isolation signal will also be sent to the feedwater isolation motor-operated valve.
This valve closes more slowly than the feedwater regulating valve.
However, initiating-automatic close of this valve on a high steam generator level represents a safety improvement for the case of a high-level signal combined with a postulated mechanical failure of a feedwater regulating valve.
Based on CYAPCO's commitment to include automatic closure of the feedwater isolation motor-operated valves in this project, CYAPC0 hereby proposes to close out SEP Topic IV-2 and ISAP Topic 1.54.
By letter dated April 23, 1992,'" the Staff indicated that the feedwater control system upgrade submittal should include a description of the isolation (3)
A.
B. Wang letter to E. J. Mroczka, " Safety Evaluation of Northeast Utilities' Topical Report 151, 'Haddam Neck Non-LOCA Transient Analysis' (TAC No. 61990)," dated October 18, 1988.
(4)
U.S. Nuclear Regulatory Commission letter to.J. F. Opeka, " Isolation of Reactor Protection System from Non-safety Systems," dated April 23, 1992.
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4 U.S. Nuclear Regulatnry Commission B14342/Page 4 January _29, 1993 devices selected.
Qualified isolation will be-provided in accordance with-.
Specifically, CYAPC0 has elected to use Foxboro Spec 200, N-2AO-VAI for analog isolation and N-2AO-L2C-R for digital: isolation.
Description of Proposed Chanae to Technical Soecification The modifications being made to the feedwater control system necessitate-changes to the Haddam Neck Plant technical soecifications to reflect changes 4
in system design, duration, and testing. -These proposed changes will rov _ise -
Technical Specification Tables 3. 3-1, 4. 3 -1,
- 3. 3_- 2, 3. 3-3, 4. 3 -2, and Bases 3/4.3.1 and 3/4.3.2.
These technical specifications govern reactor trip system instrumentation, surveillance. requirements 'and bases, and engineered-4 safety
- features, actuation system instrumentation, trip setpoints, surveillance requirements and bases.
Each change is delineated below:
functional unit.9 (steam generator; water level-low I.
Table 3.3-1 coincident with steam /feedwater flow mismatch)
The total number of channels is-being changed from one/ steam generator level to three/ steam generator _ level and one/ steam /feedwater flow mismatch in each steam generator to two/ steam /feedwater flow mismatch in each steam generator.
1 The channels to trip 'is being/feedwater flow mismatch in the same loop. change level coincident with one/ steam The minimum channels operable for steam generator level is being changed from one to two/ steam generator. -Action No. 5 is being deleted.
Action No. 6 is being invoked.
i These changes support the installation of three narrow-range level, two steam flow and two feedwater flow transmitters per steam-generator._ as opposed to the existing design of one narrow-range level, one-steam flow, and one feedwater flow transmitter per steam generator.
With ;three 4:
narrow-range level transmitters-per steam generator, Action-No. 5 is -no longer applicable. -The action requirements are met by Action No. 6.
II. Table 4.3 functional unit 8 (steam flow--high) and unit 9_ (steam generator water level-low coincident with steam /feedwater flow mismatch)
. The analog channel operational test surveillance requirements are being changed from refueling frequency to every six weeks to reflect on-line testing capability..
III. Table 3.3 functional unit 6.a (feedwater isolation: wide-range steam.
generator water level--high)
~Feedwater isolation on steam generator high water level will. now be-
-generated. from the narrow-range steam generator level transmitters'.
The a
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f U.S. Nuclear Regulatory Commission B14342/Page 5 January 29, 1993 total number of channels is being changed from two/ steam generator in each operating loop to three/ steam generator in each operating loop.
The channels to activate is being changed from one/ steam generator in any-operating loop to two/ steam generator in. any operating loop.
Table notations (c) and (d) are no longer applicable to functional unit 6 a.
Table notation (c) is being deleted as iot applicable to any functional unit.
All references to 6.a are being removed from table notation (d).
With three narrow range level transmitters per steam generator, Action No. 26 is no longer applicable.
The action requirements are met by Action No. 24.
IV. Table 3.3 functional unit 6.a (steam generator water level--high)
Ihe engineered safety features actuation system instrumentation trip setpoint for feedwater isolation on steam generator high water level is being changed from s 69 percent of wide-range instrument span to s 74 percent of narrow range instrument span.
The allowable value is being changed from s 72 percent of wide-range instrument span to s 80 percent of narrow-range instrument span.
This value is conservative with regards to overfill protection and will still allow plant operations without spurious overfill protection.
V.
Table 4.3 channel functional unit 6.a (steam generator water level--
high)
The engineered safety features actuation system instrumentation turveillance requirements for feedwater isolation on steam generator high water level trip activating device. operational test is being changed from not applicable to refueling frequency.
The new circuitry allows for a complete logic test to be performed.
VI.
Bases for 3/4.3 (instrumentation), 3/4.3.1 and 3/4.3.2 reactor trip system instrumentation and engineered safety features activation system instrumentation Add the following paragraph:
The feedwater isolation circuit is considered to be operable provided either the feedwater regulating valve or the feedwater isolation motor-operated valve is available for automatic closure.
SAFETY ASSESSMENT The proposed changes to the technical specifications reflect the plant design change to modernize the feedwater control system. The technical specification changes. to; the reactor trip system instrumentation requirements reflect-the added redundancy for steam generator flow, steam generator narrow-range level, and the steam flow / feed flow mismatch reactor trip.
The frequency of reactor
1 U.S. Nuclear Regulatory Commission B14342/Page 6 January 29, 1993 trip system instrumentation surveillance requirements for the high steam flow trip and the steam flow / feed flow mismatch is being increased from once every refueling to once every six weeks to reflect the on-line testing capability.
The technical specification changes to the engineered safety features actuation system instrumentation requirements reflect the steam generator narrow-range level being utilized in place of the steam generator wide-range level for feedwater isolation function. Accordingly, this change reflects the more conservative setpoint from the narrow-range instrument and the increase in the total number of channels and the channels to actuate, as well as the minimum number of channels operable.
A surveillance requirement to perform a trip actuating device operational test of the steam generator high water-level isolation every refueling outage has been added.
The feedwater isolation change in allowable value and trip setpoint is more conservative than the current technical specification, in that it provides additional overfill protection against the excess feedwater event.
In addition, the change in the operability and surveillance requirements are more restrictive than the present requirements for both the reactor trip system and the engineered safety features actuation system instrumentation.
The existing technical specifications provide a footnote ([c] for Table 3.3-2) which clarifies that the feedwater isolation instrumentation be operable only when feedwater is in automatic control.
Deletion of this footnote from the technical specifications will not change the design function of the feedwater isolation since the automatic closure of the feedwater isolation motor-operated valve on the steam generator high-level setpoint is assured all the time regardless of the feedwater flow control mode; i.e., automatic or manual.
This proposed change will also enhance conformance with the single-failure criterion by:
(a) increasing channel and train independence and redundancy, (b) providing a design and test scheme to eliminate identified nondetectable failures, (c) reducing potential cascaded and design-basis event-failures, and (d) eliminating certain common mode failures through the use of qualified Class 1E components.
SIGNIFICANT HAZARDS CONSIDERATION CYAPC0 has reviewed the proposed changes to the technical specifications -in accordance with 10CFR50.92 and has concluded that.the changes do not -involve a significant hazards consideration.
The basis for this conclusion is that the three criteria of 10CFR59.92(c) are not compromised.
The proposed changes do not involve a significant. hazards consideration because the changes would not:
1.
Involve a significant increase in the possibility of occurrence or consequences of an accident previously analyzed.
The proposed l'
U.S. Nuclear Regulatory Commission B14342/Page 7 January 29, 1993 changes to the reactor trip system instrumentation requirements reflect the added redundancy of steam generator narrow-range level, steam and feed flow, and the reactor trip system instrumentation surveillance requirements from every refueling cycle to every six weeks to reflect the on-line testing capability.
The proposed changes to the engineered safety features actuation system instrumentation requirements reflect the steam generator narrow-range level being utilized in place of steam generator wide-range level for feedwater isolation.
The proposed changes reflect the increase in the total number of channels and the channels to actuate feedwater isolation.
The proposed changes also provide a steam generator high-level setpoint more conservative than the current wide-r 79e setpoint for the narrow-range instrumentation for feedwatt.c isolation.
The proposed changes also add a surveillance.
requirement to perform a trip actuating device operational test of feedwater isolation every refueling outage. Thus, the change in the operability and the surveillance requirements will enhance the reliability and the availability of the reactor trip system and the engineered safety features actuation system at the Haddam Neck Plant.
There is no adverse impac+ " any design basis analysis due to the change.
2.
Create the possibility of a new or different kind of accident from any accident previously analyzed.
The change in operability and the surveillance requirements in the feedwater, and steam flow instrumentation and -the feedwater isolation function will enhance the reliability and the availability for the system.
Since thore are essentially no changes in the-way the plant is operated, the potential for an unanalyzed accident is not created. No new failure modes are introduced.
3.
Involve a significant reduction in a margin of safety. The proposed changes do not have any adverse impact on the protective boundaries, The change in the operability and surveillance requirements are more t
restrictive than the present requirements for both the reactor trip system' and the engineered safety feature actuation system instrumentation.
The new narrow-range steam. generator high-level setpoint.is more conservative than the current wide-range setpoint; therefore, there is no adverse impact upon excess feedwater flow.
Since the proposed changes also do not affect the consequences of any accident previously analyzed, there is no reduction-in a margin of safety.
In summary, for the reasons identified above, CYAPC0 has concluded that continued operation of the facility in accordance~with the proposed amendment would not involve a significant hazards consideration.
I Moreover, the Commission has provided guidance concerning the application of standards in 10CFR50.92 by providing certain examples (March 6,
- 1986,
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U.S. Nuclear Regulatory Commission B14342/Page 8 January 29, 1993=
SIFR7751) of amendments that are considered not likely to invalve a significant hazards consideration.
Example (ix) addresses a repair or replacement of a major. component or system-important to safety provided two conditions are met.
The proposed changes to technical s)ecifications reflect an improved design for feedwater control, j
added redunc ancy, and on-line testing capability (not previously available).
Example (ix), Criterion (1) is met in that the repair or replacement process involves practices which have been successfully implemented at least once on similar components or systems elsewhere in the nuclear industry, or in other industries, and does not involve a significant increase in the probability or conseunces of an accident previously evaluated or create the possibility _of a new or different kind of accident from any accident previously evaluated.
Criterion (2) is met in that the repaired--or replacement component or system does not result in - a significant change in its safety function. or a
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l significant reduction in any safety limit (or limiting condition of. operation)
L associated with the component or system.-
CYAPC0 has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations.-
The proposed changes do not involve a-significant hazards consideration, nor increase the types and-amounts of effluents-that may be released off site, nor significantly increase individual or cumulative occupational. radiation exposures.
Based on the-L foregoing, CYAPC0 concludes that the proposed changes meet the criteria delineated in 10CFR51.22(c)(9) for a
categorical exclusion from the requirements of an environmental impact statement.
Revision bars are provided.in the right-hand margin to indicate a revision to text.
No revision bars are - utilized when the page is changed solely to accommodate-the shifting of text due to additions or deletions.
The Haddam Neck-Plant Nuclear Review Board has reviewed and approved the proposed changes and has concurred with the above determination.
In accordance with 10CFR50.91(b),-we are. providing the State of Connecticut-l with a copy of the. proposed amendment.
Regarding our proposed schedule for this amendn.ent, -we -request -issuance at
_your earliest convenience, but no later _than July 1, -1993, with the amendment i
effective - as of the date of issuance, to be implemented -within 30. days of issuance.
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3 Should you have any questions, please-contact us.
Very truly yours, CONNECTICUT _ YANKEE ATOMIC POWER COMPANY bs4 L
L J. F. OpakaJ_
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Executive Vice President cc:
T. T. Martin, Region I Administrator i
A. B. Wang, NRC Project Manager, Haddam Neck Plant W. J. Raymond,. Senior Resident Inspector, Haddam Neck Plant Mr. Kevin McCarthy-i Director, Radiation control Unit i
Department of. Environment' Protection-
' Hartford, CT 06116 Subscribed and sworn-to before me l
this M day of 3 Awamto, 1993 OO-l lotary Public t--
Date Commission Expires: lb 11
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