ML20212K222
| ML20212K222 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/23/1987 |
| From: | Hukill H GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20212K220 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 NUDOCS 8701290008 | |
| Download: ML20212K222 (8) | |
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s METROPOLITAN EDISON C@iPANY JERSEY CENTRAL POWER & LIGHT COMPANY AND PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION, UNIT 1 Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No.166 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station, Unit 1.
As a part of this request, proposed replacement pages for Appendix A are also included.
GPU NUCLEAR CORPORATION BY: Vice President & Director, TM Sworn and Subscribed to befo e me this d23M day of, //nd/Mt?, 1967.
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$!'AUN P. C20FN, NOTARY PU3 llc ElC0tif0M SORO, DAUPHIN COUNTY ltY COMr.!!5110N EXPlRCS JUN! 12,1939 blember, Pennsylvania Association of Notaries I
UNITED STATES OF AMERICA NUCLEAR REGULATORY CGWISSION IN THE MATTER OF DOCKET NO. 50-289 LICENSE NO. DPR-50 GPU NUCLEAR CORPORATION This is to certify that a copy of Technical Specification Change Request No.166 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with executives of Londonderry Township, Dauphin County, Pennsylvania; Dauphin County, Pennsylvania; and the Pennsylvania Department of Environmental Resources, Bureau of Radiation Protection, by deposit in the United States mail, addressed as follows:
Mr. Jay H. Kopp, Chairman Mr. Frederick S. Rice, Chairman Board of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Courthouse Middletown, PA 17057 Harrisburg, PA 17120 Mr. Thomas Gerusky, Director PA. Dept. of Environmental Resources Bureau of Radiation Protection P.O. Box 2063 Harrisburg, PA 17120 i
GPU NUCLEAR CORPORATIDN I
BY:
Vi~ e Presi dent A Director, TMI-l i
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DATE:
January 23, 1987 1
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TECHNICAL SPECIFICATION CHANGE RE0 VEST NO. 166 The licensee requests that the following page changes be made:
(a) Replace the following pages:
3-25 3-28a 4-7 3-26b 3-32 4-7a 3-27a 3-32a 4-52 3-28 4-1 4-52a (b) Add page 3-32b (c) Delete page 4-52b II.
REASON FOR CHANGE The purpose of TSCR 166 is to provide operability and surveillance requirements and other conforming changes associated with the Emergency Feedwater System (EFW) long term upgrades (NUREG 0737, II.E.1) and to
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include administrative or editorial changes to the affected T.S. pages.
III.
SAFETY EVALUATION JUSTIFYING THE CHANGE EFW long term upgrades were required by NUREG 0737, Item II.E.1.
EFW System improvements, including the Heat Sink Protection System (HSPS),
were required by the Licensing Board to be installed prior to startup following Cycle 6 Refueling (LBP-82-27,15 NRC at 747 and TMI-l Condition of Operation, 3(a) dated October 2,1985). The purpose of the modifications was to upgrade EFW to a safety grade system in order to provide increased EFW system reliability in mitigating the effects of design basis accidents when the Main Feedwater (MFW) System is not av ailable.
Since loss of MFW is an anticipated event, EFW system reliability should be high.
The HSPS provides for automatic initiation of EFW and OTSG water level control independent of the Integrated Control System (ICS). As part of the modifications to improve the reliability of the EFW System, GPUN has added redundant EFW control and block valves, and automatic EFW System initiation on OTSG low water level and high containment pressure. High containment pressure would be an indication of a main steam line break (MSLB), a main feedwater line break (MFLB) or a loss of coolant accident (LOCA).
Automatic isolation of MFW on low OTSG pressure has been added such that MFW would be isolated in the event of a loss of OTSG integrity.
This TSCR proposes T.S. changes in order to provide operability and surveillance requirements for EFW, including the HSPS, similar to that of Safety Related Systems.
k HSPS operability requirements have been incorporated into Section 3.5-1, Operational Safety Instrumentation. Allowable outage times.
specified for'the redundant trains of HSPS instrumentation in the t
proposed specification '3.5.1.9 and Table 3.5-1.D are equivalent to the allowable outage times specified.for the separate flow path " trains" of EFW as provided in T.S. 3.4.1.1.b.
HSPS has been included in Table
.3.5-1 as a separate Section (D) since, unlike the instruments listed in Section (C), EFW does not receive an ES Actuation and is not an ESAS system. Section 3.5.1 Bases have also been revised to include HSPS Bases.
The minimum number of operable channels of HSPS instruments and minimum degree of redundancy proposed are similar to that of the RPS.
Operability requirements for HSPS instrumentation assure that no single active failure will prevent the automatic initiation of EFW or the automatic isolation of Main Feedwater. The provision of automatic features, however, necessarily. increases the probability of inadvertent actuations.
Inadvertent initiation of EFW would result in an undesirable thermal cycle for the EFW nozzles and a thermal shock to the OTSG tubes, and other OTSG structural components.
But this should not result in a tube rupture or any other accident.
Inadvertent EFW initiation would not prevent the mitigation of any accident or prevent attaining Cold Shutdown conditions.
Inadvertent Main Feedwater isolation is also undesirable. However, the probability of a loss of Main Feedwater due to single active failures during HSPS operation in accordance with these proposed Technical Specifications is not i
significantly increased, since HSPS is designed to be single failure proof. Loss of Main Feedwater is an anticipated event, the consequences of which are lessened to a certain extent as a result of 1
the automatic EFW initiation features provided by the HSPS.
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Surveillance requirements for HSPS instrumentation are incorporated l,
into Table 4.1-1.
Startup and Operating range OTSG level are part of HSPS and therefore these instruments have been included with the other HSPS instrumentation under a new Item No. 51, HSPS. However, full range OTSG 1evel is not a part of HSPS. Therefore, Item No. 38, Steam Generator Water Level has been revised to include only the Full range level instruments.
Table 3.5-1, Item C.3.a and Table 4.1-1, Item 19.a, high Reactor Building Pressure instrumentation, are unaffected by this TSCR. The l
HSPS uses separate new instruments for high Reactor Building Pressure.
l Operability and surveillance requirements for the new instruments l
therefore should be included separately in the instrument tables along i
with the other HSPS instrumentation.
l Instrumentation for automatic EFW initiation on Loss of both Main Feedwater Pumps and on Loss of all RC Pumps have been included in the proposed T.S. along with other HSPS instrumentation.
As a change from the presentation in the current T.S., Table 3.5-1 Channel Operability and Redundancy for this instrumentation is shown to be not applicable (N/A) since this instrumentation does not fit the definition of a l
"cha nnel. " It should be noted that this change does not result in a change in operability or redundancy requirements since the train operability requirements of the proposed specification 3.5.1.9 preserves the operability requirements of the current T.S., as described in Note (b) of the Table.
The proposed Table 4.1-1 deletes the remark "when Tave is greater than 250*F" which refers to Surveillance Testing. This statement is not consistent with the operability requirements of T.S. 3.5.1 and therefore 'should be deleted. The operability requirements for EFW automatic initiation instrumentation included in the proposed T.S are the same as provided in the current T.S. 3.5.1.1.
In order to minimize the potential for inadvertent EFW initiation during Startups and Shutdowns, the OTSG low water level EFW initiation circuitry may be bypassed by the operator whenever indicated reactor power is less than 30%. This is because of the potential for OTSG 1evel oscillations or transients that may be experienced at low power.
This bypass must be removed, by operator action, when indicated reactor power is above 30%.~ This operational bypass is in addition to the bypass of MFW isolation on low OTSG pressure that was discussed in GPUN's letter to the NRC dated March 14, 1986. Although OTSG level oscillations and transients are not expected to occur at low power levels, the potential for inadvertent EFW initiation as a result of changes in equipment operating characteristics or personnel errors warrant bypassing the OTSG low level setpoint at low power levels.
GPUN believes that bypass of the OTSG low level actuation for automatic EFW initiation at low power is justified in order to avoid spurious EFW y
initiations. Loss of Main Feedwater events are less severe at low por r and more time is available for the operator to manually initiate EFW. 30% power was chosen as the appropriate power below which the OTSG low level actuation feature may be defeated. This power level was arrived at as a result of engineering judgement.
30% is a high enough power level above which spurious actuations would be less likely to occur.
There is no significant risk of core damage or reactor coolant system pressure boundary rupture caused by bypass of the low level
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setpoi nt. Automatic initiation of EFW is either not required or is l-accomplised by diverse signals except for a Main Feedwater line break l
in the intermediate building downstream of the feedwater regulating valves.
Even for this break, the accident acceptance criteria for this accident can be met without automatic EFW initiation because RCS pressure will not exceed the code allowable limit of 2750 psig, and no departure from nucleate boiling (DNB) will occur for a Main Feedwater line break initiated at this power level.
Analyses have previously been reviewed by the NRC staff which show that operator action at 20 minutes following a total loss of feedwater transient will prevent core uncovery.
Since this accident would be initiated at a power level of no more than 30% ' ower, the operator p
would have a substantial period of time to manually initiate EFW or restore Main Feedwater flow. '
T This TSCR also includes editorial changes to the instrumentation tables that do not involve HSPS instrumentation.
Note (g) in Table 3.5-1 for Item C.3.a should be deleted as well as the asterisk in Table 4.1-1, Item No. 39. These deletions are editorial in nature since the statements that are being deleted were applicable only for the previous cycle of operation (Cycle 5) as stated in the entries that are being deleted. Also, Amendment No. 78 lef t a blank in Table 4.1-1, Item No. 47a under the Test column for the Pressurizer Code-Safety Valves and PORV Tailpipe Flow Monitors.
NRC's safety evaluation for Amendment No. 78 lists the surveillance requirements for this item which appear in Item No. 47a and no mention is made of any test that would be applicable to this item for the column which appears blank. Since such a test as would be indicated in the Test column is not appropriate for this item, "N/A" has been inserted to fill the blank. This clarification is needed to remove the ambiguity of the blank table entry.
"N/A" appears elsewhere in Table 4.1-1 where an entry is not applicable.
In addition to the T.S. changes which relate to instrumentation, this TSCR includes other changes related to the EFW system.
These changes are needed in order to reflect the upgrade of EFW to Safety Grade and to improve the clarity of these specifications.
1)
Sections 3.4.1.1.a and b provide sufficient action statements for EFW system inoperability.
Section 3.4.1.1.d provides sufficient action statements for CST inoperability. Therefore, Section 3.0.1 action statements are not necessary and the references to Section 3.0.1 in Sections 3.4.1.1.a and 3.4.1.1.d have been deleted.
2)
GPUN proposes to delete the flowpath operability definition associated with Sections 3.4.1.1.b and 4.9 l.2.
The EFW System has been upgraded to Safety Grade.
EFW System operability is required in order to meet the conditions assumed in the safety analysis taking into account a single active failure. Therefore, this definition does not appear to be applicable to the Safety Grade EFW System and should be deleted.
3)
Section 3.4 Bases have been revised to clarify that either two motor-driven EFW pumps or the steam-driven pump are sufficient for,
decay heat removal immediately following any shutdown.
4)
Requirements for functional surveillance testing each refueling have been added to Specification 4.9.1.
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5)
The manual control valve station (HIC-849/850) which was part of l
the interim system design has been replaced as part of the long term modifications during the Cycle 6 refueling outage.
l Therefore, testing requirements related to this temporary valve l
station have been deleted.
6)
Full-stroke testing of the EFW control valves is required by l
T.S. 4.2.2.
Therefore, the redundant requirement for full-stroke testing of the EFW control valves in T.S. 4.9.1 is being deleted.
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7)
Table 4.9-1 is being deleted. The requirement to verify the is position of EFW flowpath valves every 31 days is adequately stated I
in T.S. 4.9.1.3 and implemented by SP 1300-3F and SP 1300-3G.
Table 4.9-1 specifies a level of detail that need not be included in the T.S. and should be deleted.
8)
Section 4.9 Bases have been revised to delete statements which GPUN believes are inappropriate for inclusion in the Bases. The sentence which refers to a B&W EFW Reliability Study as having demonstrated the 31 day test frequency to assure an appropriate level of reliability may be incorrect in that some other test frequency may provide an equivalent or higher level of-i reliability. Therefore, this sentence should be deleted.
EFW minimum flow as described in the bases of the current T.S. are incorrect. Minimum EFW flow requirements have been demonstrated through analysis as described in GPUN's letter of December 9,1983 and reflected in the NRC's SER dated April 27, 1984.
In accordance with 10CFR50.55a, the ability of the EFW pumps to meet accident flow requirements is demonstrated through the Inservice i
Testing Program. Therefore,' statement of the analyses are beyond the level of detail required for these Bases, and the incorrect bases statements should be deleted.
The purpose of the EFW system modifications is to increase the reliability of the EFW System in its capacity to withstand a design basis event and a single active failure and sti11' function to supply a heat removal path to allow cooldown of the RCS to a safe shutdown condition.
Operation in accordance with these specifications will not 4
degrade the reliability or capability of the EFW System.
IV.
NO SIGNIFICANT HAZARDS CONSIDERATION i
GPUN has determined that this TSCR poses no significant hazards as defined by NRC -regulations in 10 CFR 50.92.
1.
Operations of the facility in accordance with the proposed amendment would not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated. The proposed technical specifications include additional operability and surveillance requirements with appropriate bypass capability to minimize any increase in the l
potential for inadvertent MFW isolation or inadvertent EFW l
initiation that may result from the addition of the new automatic features. These features were added in conjunction with other j
design changes the purpose of which was to increase the reliability of the EFW system and thereby reduce the probability i
of occurrence or consequences of accidents (10 CFR 50.92(c)(1)).
2.
Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different l
kind of accident from any accident previously evaluated. Loss of l
MFW and any event which might result from inadvertent EFW System initiation or failure of the equipment installed as part of the t
EFW System upgrades are all events which have been analyzed.
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3.
Operation of the facility in accordance with the proposed amendment would not twolve a significant reduction in a margin of safe ty. The purpose of the EFW system upgrades was to increase the EFW System reliability. Therefore, operation in accordance with these specifications would serve to increase the overall margin of safety for plant operation.
(10 CFR5 0.92(c)(1))
The Commission has provided guidelines pertaining to the application of the three standards by listing specific examples in 48FR14870 of technical specification changes that do not pose significant hazard considerations. The proposed amendment is considered to be the same category as the following examples of amendments that are not likely to isolve significant hazard considerations:
(1) A purely administrative change to achieve consistency, correct errcrs, change nomenclature and improve clarity.
(ii) A change that constitutes an additional limitation, restriction, or control not presently included in the technical specifications.
(v) Upon satisfactory completion of construction in connection with an operating facility, a relief granted from an operating restriction that was imposed because the construction was not yet completed satisfactorily. This is intended to isolve only restrictions where it is justified that construction has been completed satisfactorily.
Therefore, the three standards of 10 CFR 50.92(c) are satisfied and operation of the facility in accordance with the proposed amendment isolves no significant hazard consideration.
V.
IMPLEMENTATION It is requested that the amendment authorizing this change become effective upon issuance VI.
AMENDMENT FEE (10 CFR 170.21) l Pursuant to the provisions of 10 CFR 170.21, attached is a check for
$150.00.
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