ML18093A270

Application for Amends to Licenses DPR-70 & DPR-75,changing Reactor Coolant Pump Breaker Position Trip Logic from 1 Out of 4 Logic Above 36% Power Level (Permissible P-8) to 2 Out of 4 Logic Above 11% (P-7 Permissive).Fee Paid

ML18093A270
Person / Time
Site:	Salem
Issue date:	07/24/1987
From:	Corbin McNeil Public Service Enterprise Group
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML18093A271	List: ML18093A270 ML18093A272
References
LCR-87-11, NLR-N87138, NUDOCS 8708030377
Download: ML18093A270 (8)

Text

Public Service Electric and Gas Company Corbin A. McNeil!, Jr. Public Service Electric and Gas Company P.O. Box236, Hancocks Bridge, NJ 08038 609 339-4800 Senior Vice President -

Nuclear July 24, 1987 NLR-N87138 LCR 87-11 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

REQUEST FOR LICENSE AMENDMENT FACILITY OPERATING LICENSE DPR-70 AND DPR-75 SALEM GENERATING STATION - UNIT NOS. 1 AND 2 In accordance with the Atomic Energy Act of 1954, as amended, and the regulations thereunder, we hereby transmit our request for amendment and our analyses of the changes to Facility Operating Licenses DPR-70 and DPR-75 for the Salem Generating Station Unit Nos. 1 and 2, respectively.

The amendment request consists of changing the Reactor Coolant Pump Breaker Position Trip logic from a 1 out of 4 reactor trip logic above 36% (permissive P-8) to a 2 out of 4 reactor trip logic above power level 11% (P-7 permissive). This change will remove a potential source of single failure unit trips and provide a reduction in challenges to the Reactor Protection System. This modification is provided as a standard feature or as a retrofit for all Westinghouse PWR units. Note that Technical Specification pages 3/4 3-7 for both Salem Unit 1 and 2 included in this submittal were also affected by a previous submittal (LCR 87-09 dated July 2, 1987). The pages submitted in this document include the previously requested changes as well as those change requested by this submittal. These modifications are currently scheduled for installation during the Unit 1 Cycle 7 outage (October 1987) and for the Unit 2 Cycle 4 outage (April 1988). Therefore, approval by October 1987 is requested.

Enclosed is a check in the amount of $150.00 as required by 10CFR 170.21.

Pursuant to the requirements of 10CFR50.91, a copy of this request for amendment has been sent to the State of New Jersey as indicated below. This submittal consists of one (1) signed original and thirty-seven (37) copies. ~

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Document Control Desk 2 July 24, 1987 Should there by any questions regarding this matter, please feel free to contact us.

Sincerely, Attachment C Mr. D. c. Fischer USNRC Licensing Project Manager Mr. T. J. Kenny USNRC Senior Resident Inspector Mr. w. T. Russell, Administrator USNRC Region I Mr. D. M. Scott, Chief Bureau of Nuclear Engineering Department of Environmental Protection 380 Scotch Road Trenton, NJ 08628

Ref: LCR 87-11 STATE OF NEW JERSEY )

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COUNTY OF SALEM )

Corbin A. McNeill, Jr., being duly sworn according to law deposes and says:

I am Senior Vice President of Public Service Electric and Gas Company, and as such, I find the matters set forth in our letter dated July 24, 1987 , concerning Facility Operating License DPR-70 and DPR-75 for Salem Generating Station, is true to the best of my knowledge, information and belief *

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LICENSE CHANGE REQUEST DCR-1EC/2EC-2245 I. DESCRIPTION OF THE CHANGE Revise the Technical Specification Bases, Section 2.2.1 Limiting Safety System Settings, "Reactor Coolant Pump Breaker Position Trip", and Table 3.3-1, Reactor Trip System Instrumentation, Functional Unit 20 "Reactor Coolant Pump Breaker Position Trip" to change the (1/4) RCP breaker open position anticipatory reactor trip logic above power level P-8 into (2/4) RCP breaker open position reactor trip logic above power level P-7. (see Attachment 1) This modification removes a potential source of single failure unit trips and provides a reduction in challenges (e.g. spurious trips) to the Reactor Protection System. This modification is provided as a standard feature or as a retrofit option for all Westinghouse PWR units. All domestic Westinghouse PWR units supplied subsequent to Three Mile Island (TMI) incorporate this change.

II. REASON FOR THE CHANGE The (1/4) RCP breaker open position reactor trip is primarily an anticipatory trip. If an RCP breaker opens, the low flow reactor trip setpoint is typically reached two to three seconds after the RCP begins coasting down. Thus the RCP breaker open position logic will trip the reactor in anticipation of the low flow reactor trip. Low flow in any one of four reactor coolant loops above the P-8 power level setpoint (greater than or equal to 36% of rated thermal power) and low flow in any two of four reactor coolant loops between power levels P-7 (greater than or equal to 11%

of rated thermal power) and P-8 will trip the reactor.

The low flow reactor trip is blocked below power level P-7.

The existing RCP breaker open position reactor trip logic is identical to the low flow reactor trip logic (i.e. (1/4) above P-8, (2/4) between P-7 and P-8, blocked below P-7).

Licensee Event Report LER-85-022 dated November 6, 1985 describes the circumstances which led to the unit trip/Safety Injection Event at Salem, Unit 2 on October 7, 1985. In that instance, an RCP breaker open position reactor trip was actuated by deenergizing the corresponding "RCP breaker tripped" interposing relay in the SSPS logic cabinet when its Vital Instrument Bus power source experienced a momentary ground fault.

This same fault condition deenergized the steam line loops 22 and 23 pressure protection instrument loop power supplies which fulfilled the (2/4) permissive logic for actuation of the Safety Injection System.

This momentary fault on a single Vital Instrument Bus panelboard resulted in a Unit Trip/Safety Injection by the "simulated" opening of an RCP breaker.

The existing and proposed low flow and RCP breaker position reactor trip logic is respectively shown in Figures 1 and 2 of Attachment 2.

III. JUSTIFICATION FOR THE CHANGE The (1/4) RCP breaker open position reactor trip above power level P-8 is not required for safety but is instead provided as a redundant anticipatory trip for an impending loss of reactor coolant flow to enhance the overall reliability of the Reactor Protection System. In the event of a single loop loss of flow, the low flow reactor trip is the design protection.

The proposed change in Reactor Protection System logic modifies the (1/4) RCP breaker open position reactor trip above power level P-8 into (2/4) RCP breakers open above power level P-7 (i.e. two RCP breakers open will be required at all power levels above P-7 to initiate the RCP breaker position reactor trip). If a single RCP is coasting down, then a reactor trip will occur on low flow. UFSAR Section 15.3.4.1 shows that adequate protection is provided by the low flow reactor trip logic for low primary coolant flow in a single loop.

This latter logic is a backup to the RCP power supply bus undervoltage and underfrequency reactor trip logic. UFSAR Section 15.3.4.4 concludes that the minimum DNBR will remain above the limiting value of 1.30 during a transient in a loss of forced reactor coolant flow (one or more loops). The deletion of the (1/4) RCP breaker open position reactor trip above

power level P-8 affects only the coincidence logic of the Reactor Protection System and does not degrade its performance or conformance to system functional requirements. This change provides a reduction in challenges (e.g. spurious trips) to the Reactor Protection System while retaining diversity and redundancy of the Reactor Protection System for the loss of flow in a single loop.

IV. SIGNIFICANT HAZARDS CONSIDERATION A. The proposed changes to the Salem Technical Specifications do not significantly increase the probability or consequences of a previously evaluated accident. UFSAR Section 15.2.5, "Partial Loss of Forced Reactor Coolant Flow" describes the low primary coolant flow reactor trip as providing the required protection against a partial loss of coolant flow accident. Three redundant flow channels are provided in each loop for the low flow reactor trip. Above power level P-8, the loss of flow in any one loop, as sensed by two of the three channels will actuate the low flow reactor trip.

The existing (1/4) RCP breaker open position anticipatory reactor trip functions to backup the low flow reactor trip. This same analysis assumes a worst case scenario (e.g. four loops initially operating, two pumps coasting down). UFSAR Section 15.2.5.5 concludes that the DNBR will not decrease below the limiting balue of 1.30 at any time during such a transient thus the core safety limit is not breeched. The anticipatory time delay (e.g. the time difference between the initiations of an RCP breaker open position reactor trip and a low primary coolant flow reactor trip) is not significant since the original transient analysis is based on low primary coolant flow. Therefore, the UFSAR Section 15.2.5.5 "Conclusions" bounds the consequences of deleting the (1/4) RCP breaker open position anticipatory reactor trip above power level P-8 and hence the proposed modification does not increase the probability or consequences of this previously evaluated accident scenario.

B. The proposed changes to the Salem Technical Specifications do not create the possibility of a new or different kind of accident than any previously evaluated. In the event of a single loop loss of flow, the low primary coolant flow reactor trip is the design protection and it meets the design requirement of maintaining the minimum DNBR above the 1.30 limiting value. If credit is not taken for the low primary coolant flow reactor trip, then a reactor trip on either overtemperature or overpower Delta-T will terminate the accident before DNB occurs in the core (UFSAR Section 7.2.3.1). Westinghouse analysis indicates that the hot spot clad temperature (on the inner clad surface) will remain well below the melting point (reference Attachment 2). For all non-LOCA safety analyses, the logic change which results in the deletion of a reactor trip on a single RCP breaker open position above power level P-8 is acceptable since UFSAR Sections 15.2.5.5 and 15.3.4.4 "Conclusions" bound the consequences. The diversity and redundancy of the Reactor Protection System are maintained for a single loop loss of flow.

c. The proposed changes to the Salem GS Technical Specifications do not involve a significant reduction in the margin of safety. The RCP breaker position reactor trip is modified to actuate on the opening of any two breakers above the P-7 interlock setpoint. This modification only affects the coincidence logic of the Solid State Protection System (SSPS). The existing (2/4) coincident reactor trip logic functions (e.g. the opening of an RCP breaker in any one loop coincident with low flow in a second loop or low flow in any two loops simultaneously) will remain unchanged. This specific modification has undergone Westinghouse analysis and review in accordance with the general requirements of 10CFR50 Appendix B and the specific requirements of IEEE-279-1971 and does not degrade either its performance or conformance to system design requirements. Two independent RCP breaker open position interlocks are provided to the SSPS logic cabinets to develop the RCP breaker position reactor trip logic (independent interlocks for trains "A" and "B"). The single failure criterion requirement for independence and redundancy of the (2/4) RCP breaker open position reactor trip above power level P-7 is not changed as a result of this modification thus UFSAR Section 7.2.2.2 "Design Basis for Protection Circuits" is

not changed and the existing margins of safety are retained. Based on the preceding discussions in IV. A, B and c, PSE&G concludes that the proposed change to the Salem Technical Specifications does not involve a significant hazards consideration.

(Reference Attachments 3 and 4).

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