ML18093A211

Application for Amends to Licenses DPR-70 & DPR-75,changing Reactor Trip Block W/Turbine Trip from P-7 (11% Power) Permissive Up to P-9 (50% Power) Permissive.Fee Paid

ML18093A211
Person / Time
Site:	Salem
Issue date:	07/02/1987
From:	MCNEILL C A Public Service Enterprise Group
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML18093A212	List: ML18093A211 ML18093A214
References
LCR-87-09, LCR-87-9, NLR-N87124, NUDOCS 8707110017
Download: ML18093A211 (17)

Text

I I I Public Service Electric and Gas Company

• Corbin A. McNeill, Jr. Senior Vice President

-Nuclear Public Service Electric and Gas Company P.O. Box236, Hancocks Bridge, NJ 08038 609 339-4800 July 2, 1987 NLR-N87124 LCR 87-09 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 trip block with a turbine trip from the P-7 (11% power) permissive up to the P-9 (50% power) permissive.

This will allow the Salem units to sustain a turbine trip without causing a reactor trip up to 50% of rated thermal power. This change allows operation similar to that licensed on numerous other Westinghouse plants. These changes can be incorporated during any plant outage and currently scheduled for the cycle 4 outage for Unit 2 (April 1988) and for P-8 permissive modifications during the cycle 7 outage (September 1987) and P-9 permissive modifi6ations during the cycle 8 outage (April.1989) for Unit 1. Therefore, approval by September 1987 is requested.

Enclosed is a check in the amount of $150.00 as required by lOCFR 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. 8707110017 870702 -------, PDR ADOCK 05000272 I P PDR A *. .;ff.lo

1. tl'l1

• Document Control Desk 2 7-2-87 Should there be any questions regarding this matter, please feel free to contact us. Sincerely, Enclosures C Mr. D. C. Fischer USNRC Licensing Project Manager Mr

• T

• J

• Kenny USNRC Senior Resident Inspector Mr. w. T. Russell, Administrator USNRC Region 1 Mr. D. M. Scott, Chief Bureau of Nuclear Engineering Department of Environmental Protection 380 Scotch Road Trenton, NJ 08628 STATE OF NEW JERSEY COUNTY OF SALEM ) ) ) Ref: LCR 87-09 SS. 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 2, 1987 , concerning Facility Operating Licenses DPR-70 and DPR-75 for Salem Generating Station, is true to the best of my knowledge, information and belief. Subscr_i!?,Zd and this J

• day of , . . . -.Z me 1987 LARAINE Y. BEARD Notary Public of New Jersey My Commission Expires May 1, 1991. My Commission expires on -----------------

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• ATTACHMENT 1 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS SALEM GENERATING STATION, UNIT NOS. 1 & 2 FACILITY OPERATING LICENSE DPR-70 & DPR-75 DOCKET NOS. 50-272 & 50-311 I. Desci-i pti on of the Change

• LCR 87-09 Revise Technical Specification Bases Section 2.2. 1, Tui-bine Trip, and Table 3.3-1, Reactor Trip System Insti-umentation, to indicate that the P-9, rather that the P-7, Permissive Setpoint will defeat the automatic block or a reactor ti-ip on a turbine trip when 2 of 4 Power Range Neutron Flux Channels are greatei-than or equal to 50%, rathei-than 11%, of RATED THERMAL POWER. This change will permit continued reactor operation following turbine trips, provided that reactor power is no greater than 50% of RATED THERMAL POWER. This will enable the plant to either restart the turbine for trips which are readily correctable or to commence an orderly reactor shutdown.

II. Reason for the Change The basic design philosophy for Westinghouse Pressurized Hater Reactors <PWRs> has included a i-eactor trip following a turbine trip whenever the plant is operating above some nominal power level. This reactor trip is not required but rather exists as an anticipatory trip to prevent a trip associated with excessive reactor coolant temperature and pressure rises and to enhance the overall reliability of the Reactor Protection System <RPS>: Currently, the design of Salem Generating Station, both Unit Nos. 1 and 2, includes a reactor trip when reactor power is above 11% of RATED THERMAL POWER if a major load loss results from a loss of external electrical load or a turbine trip. For either case offsite power is available for the continued operation of vital plant components.

The SGS Turbine Bypass System <see Updated Final Safety Analysis Report <UFSAR> Section 10.4.4.1>

provides the capability to dump up to 40% of full load steam flow dii-ectly to the condenser which enables the plant to accept a step load decrease of 50% of full load from full load without a reactor trip <the remaining 10% of the 50% is an inherent capability of the Nuclear Steam Supply System (NSSS> to accept a 10% step load change>. Therefore, to trip the plant when operating below 50% of RATED THERMAL POWER upon an electrical loss or turbine trip is unnecessary if the cause or the turbine trip is readily correctable.

Hence, this License Change Request CLCR> has been developed in order to increase plant availability by significantly Page 1

• reducing the down time required to restart the plant after an unnecessary plant trip, i.e. currently reactor trips are required tor turbine trips when operating at greater than 11% or RATED THERMAL POWER but are only necessary when operating at greater than 50% or RATED THERMAL POWER. PSE&G has obtained only enough hardware in order to install the P-9 set point on SGS Unit No. 2. Hence, SGS Unit No. 1 will utilize the P-8 setpoint Cgreater than or equal to 36% of RATED THERMAL POWER> until such time as the. P-9 is available and installed.

In order to eliminate the need for a second LCR for SGS Unit No. 1 Ci.e. one for P-8 and another for P-9>, the justification and significant hazards analysis provided ror this LCR have been developed assuming the P-9 setpoints will be installed and the LCR discusses the use or a setpoint 50% or RATED THERMAL However, Subparagraph IV.B. 3 below discusses the use or an interim setpoint at 36% ot RATED THERMAL POWER CP-8> and the SGS Unit No. 1 Technical Specification revisions contained in Attachment 3 contain provisions for use or the P-8 setpoint until such time as the P-9 setpoint is installed and available.

III. Justification tor the Change A summary discussion is warranted regarding the design for tripping the plant in the event or a loss or external electrical load or a turbine when operating above some nominal power level. Currently this level is set at 11 % or RATED THERMAL POWER and is known as the P-7 permissive setp6int.

The basis for this setpoint is derived from the standard Westinghouse Technical Specifications.

For the loss or external electrical load, the turbine is automatically tripped when both generator output breakers trip open. UFSAR Section 10.2.2.4 identifies a variety or conditions under which the turbine is tripped due to station events. Once the turbine is tripped and the P-7 permissive setpoint is satisfied

<i.e. at greater than 11% or RATED THERMAL POWER>, the reactor would be tripped on signals that the turbine auto stop emergency trip fluid pressure fell below 45 psig or the tour turbine stop valves were less than 85% full open -the turbine stop valves close rapidly Ctypically within 0.1 seconds> on loss or trip-fluid pressure.

Upon closure or the tour stop valves, steam flow to the turbine is stopped which results in an almost immediate rise in secondary system temperature and pressure.

This *increase results in reduce4 beat transfer rate in the steam generator, causes the reactor coolant to expand, creates a pressurizer insurge and causes the reactor coolant system pressure to rise. These conditions would create a situation in which the reactor is tripped on high pressurizer pressure, overtemperature delta-T, or low-low Page 2 ------

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• steam generator level, if the turbine trip had not already tripped the plant. If the automatic steam dump system is operating properly, up to 40% of full load steam flow C50% of RATED THERMAL POWER> can be bypassed to the condensor without a primary system transient causing a reactor scram. The secondary-to-primary system scenerio described above would not occur even if the currently design P-7 setpoint failed to function.

Hence at power levels above 11% but below 50% of RATED THERMAL POHER, the anticipatory reactor trip upon turbine trip is unnecessary.

IV. Significant Hazards Consideration In order to develop the proposed change fully, PSE&G has evaluated a variety of plant scenarios and previously completed accident analyses to assess the impact or increasing the turbine-reactor trip setpoint to 50% from 11% of RATED THERMAL POWER. The scenerios include: i. UFSAR Section 15.2. 5 discusses the plant response to a partial loss of forced reactor coolant flow. The accident analysis was re-evaluated in light or the revised setpoint, including the impact on the 30 second turbine-generator motoring feature. The results are presented in Item A.1 below. ii. UFSAR Section 15. 2. 7 discuss.es the plant response to a loss or external load and/or turbine trip. This accident analysis was re-evaluated in light of the revised setpoint and the results are presented in Item A.2 below. iii. UFSAR Section 15.3.4 discusses the plant response to a complete loss of forced reactor coolant flow. This accident analysis was re-evaluated in light of the revised setpoint and the results are presented in Item A.3 below. iv. System> operate POWER. Turbine THERMAL in Item As discussed in Paragraph III above, the operation of the steam dump system <i.e. the Turbine Bypass was assumed in order for the plant to continue to following a turbine trip below 50% ot RATED THERMAL Therefore, the plant response to a failure or the Bypass System at between 11% and 50% or RATED POWER was evaluated and the results are presented B. 1 below. v. A study of the potential tor increased pressurizer PORV opening resulting from a turbine trip without a reactor trip below 50% of RATED THERMAL POWER was completed by Westinghouse for PSB&G. The results or this Page 3

• transient analysis are summarized in Item B.2 below and further details are provided in Attachment

2. vi. In the case or SGS Unit No. 1, the use of the P-9 setpoint is restricted until the receipt and installation or hardware required to actually change the setpoint <see the discussion contained in Paragraph II above>. Until such time, Unit No. 1 intends to operate with the P-8 setpoint as the turbine-reactor trip permissive.

Therefore, the plant response to the use or a 36% or RATED THERMAL POWER setpoint was evaluated and the results are presented in Item B. 3 below. The proposed changes to the SGS Technical Specifications:

A. Do not involve a significant increase in the probability or consequences or an accident previously evaluated.

1. UFSAR Section 15.2.5 contains an analysis or the Condition II Partial Loss of Forced Reactor Coolant Flow. The accident scenario evaluates a fault in the power supply to a reactor coolant pump which can result in a partial loss or coolant flow. Ir the reactor is at power at the time or the accident, the immediate effect or loss or coolant flow is a rapid increase in the reactor coolant temperature.

The accident analysis that the low reactor coolant flow signal is available to trip the reactor. _Above 36% of RATED THERMAL POWER <i.e. the P-8 setpoint>, low flow in any one or the four loops will actuate a reactor trip, while between 11% and 36% of RATED THERMAL POWER Ci.e. between P-7 and P-8> low flow in any two or rour_loops will actuate a reactor trip. A reactor trip signal from the reactor coolant pump circuit breaker open position is provided as an anticipatory trip signal which serves to backup the low flow signals. Normal power for each or the reactor coolant pumps is supplied through separate busses from a transformer connected to the generator.

Following any turbine trip where there are no electrical faults which require tripping the generator from the network, the generator remains connected to the network for approximately 30 seconds. The reactor coolant pumps remain connected to the generator thus ensuring full flow for approximately 30 seconds after the *reactor trip before any transfer is made. ffhen the generator trip occurs, the busses are automatically transferred to a transformer supplied from external power lines. This feature is known as turbine-generator motoring and is provided so that full reactor coolant flow is maintained to remove reactor core heat durinq Condition II overpower transients

<UFSAR Section 15.2. 5) and to prevent any pump overspeed conditions

<see UFSAR Section 5.5.1.3.9).

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• If the entire 30 second time delay rails or if an electrical fault exists such that the generator is immediately tripped from the network, a fast bus transfer to orrsite power will be initiated by the generator trip signal. This transfer ensures reactor coolant flow by transferring the reactor coolant pumps within six to ten cycles. The RPS initiates a reactor trip as protection for overpower event the turbine protection system has a mechanical overspeed trip at 1101 or turbine speed and the turbine control system and intercept valves limit the overspeed to 120% or turbine speed to prevent the or a pump overspeed condition.

Changing the turbine-reactor trip setpoint from 11% to 50% or RATED THERMAL POWER will not affect the low reactor coolant loop flow or reactor coolant pump circuit breaker open position signals. The change removes the defeat or the automatic block or reactor trip upon turbine trip from P-7, which is an independent trip signal from the two signals provided for protection against a partial loss or coolant flow. Above 50% of RATED THERMAL POWER, the 30 second turbine-generator motoring feature is still available as the P-9 setpoint only serves to trip the reactor upon a turbine trip. The 30 second time delay associated with the generator decoupling from the network is an independent feature from the turbine trip signals and even if this feature fails upon turbine sufficient RPS signals are available as described above to mitigate an overpower transient.

As a result, the UPSAR Section 15.2.5 accident analysis bounds the consequences of a turbine trip event below 50% of RATED THERMAL POWER with or without turbine-generator motoring and hence the proposed changes do not increase the possibility or consequences or this previously evaluated accident." A discussiort of a failure or the transfer following the turbine-generator motoring 30 second time limit is discussed in Items A. 2 and A.3 below and the consequences of a failure of the steam dump is discussed in Item B.1 below. 2. UFSAR Section 15.2.7 contains an analysis of the Condition II Loss of External Electrical Load and/or Turbine Trip. The description of the accident sequence identifies that the reactor would be tripped directly from a signal derived from the turbine auto stop oil pressure and/or the turbine stop valves µnless below 11% or RATED THERMAL POWER, that power level associated with the P-7 permissive setpoint.

The analysis however contains

• several conservative assumptions including:

<1> a complete loss or steam load at 102% or full power, <2> no direct reactor trip upon turbine trip, and <3> no credit for steam dump or steam generator power operated relief.valves.

Four accident evaluations were completed using the LOFTRAH digital computer program assuming beginning of life Page 5

• conditions for minimum moderator reactivity feedback and end or life conditions for maximum moderator reactivity feedback both of which were completed assuming credit for the effect or pressurizer spray and power operated relief valves <PORVs> as well as assuming no credit for pressurizer spray and PORVs. These four cases have been evaluated using the P-9 setpoint <nominal 50% power>. i. The analyses performed without pressure control, both minimum and maximum reactivity feedback cases, indicate that a reactor trip on high pressurizer pressure will occur within 6-7 seconds if the event ie initiated from full power, and within about 12-17 seconds if the event is started from 50% power. The power and temperature conditions which exist at full power are more limiting than at any partial power with respect to the minimum DNBR reached during the transient

<in fact, the DNBR will increase throughout the transient).

The pressurizer safety valve setpoint will be reached for both the full and partial power cases which assume no pressure control with the pressurizer PORVs or sprays. However, the event initiated from partial powsr will turn around raster due to the lower initial power and temperatures in addition to the lower amount or stored energy in the fuel. For both the full power and partial power evaluations, the reactor trip occurs long before the fast bus transfer is attempted, thus the loss of flow which may occur due to the fast bus transfer failure after 30 seconds or turbine-generator motoring will have no effect on the transient for either the full or partial power cases. ii. The analysi e performed at par ti al power (50%> with pressure control from the pressurizer PORVS and sprays, for both the minimum and maximum moderator reactivity feedback cases, indicate that the reactor may not trip until an undervoltage or low flow setpoint is reached after failure or the fast bus transfer 30 seconds into the event. The power, temperature and pressure conditions which exist at the time. or the loss of flow for the partial power cases are much less severe with respect to minimum DNBR than those conditions which exist tor the complete lose or flow event <UFSAR Section 15.3.4>. The largest benefit is due to the lower initial power level at time or the loss of flow. For the minimum feedback case, the reactor will essentially remain at the 50% pow.er level unti 1 the reactor trip occurs on the reactor coolant pump undervoltaoe or low flow signal. The reactor coolant average temperature increases for the partial power loss or load event; however, the RCS average temperature will not significantly increase above the nominal full power RCS average temperature and thus will not offset the DNB benefit from the large difference in power level. The power transient for the maximum reactivity Page 6 feedback case will steadily decrease from the initial power level due to the beatup and moderator feedback effect. The RCS average temperature for this case will increase, but not above the nominal full power RCS average temperature, so the resultant minimum DHBR will be greater than the minimum DNBR for the loss or flow event <UFSAR Section 15. 3. 4>. Therefore, the FSAR full power complete loss or flow event bounds the partial power cases with respect to the minimum DNBR reached during the transient.

With the PORVS and sprays available for pressure control, both the full and partial power cases will show a pressure increase on the primary side to the PORV setpoint.

After reactor trip, the pressure will decrease throughout the remainder or the transient.

The cases without pressure control are always more limiting with respect to peak pressures and, as stated above, the UFSAR full power loss of load event will bound any partial power event with respect to peak pressure.

Finally, the proposed change would increase the turbine trip to 50% or RATED THERMAL POWER and hence the reactor would not be tripped above the 11% RATED THERMAL POWER level identified in the accident description.

However, since the LOFTRAN computer program did not take credit for the direct turbine-reactor trip, whether such a trip takes place or not does not affect the results or this accident analysis.

Hence changing the turbine-reactor trip setpoint from 11% to 50% or RATED THERMAL POWER has no bearing on the results or the loss or external electrical load and/or turbine trip. As a result, the UFSAR Section 15.2.7 accident analysis bounds the consequences of a turbine trip event below 50% or RATED THERMAL POWER with or without a subsequent reactor trip and hence the proposed changes do not increase the probability or consequences or this previously evaluated accident.

3. UFSAR Section 15.3.4 contains an analysis or the Condition III Complete Loss of Forced Reactor Coolant Flow. This accident sequence evaluates the effects of a complete loss or forced reactor coolant flow from a loss of electrical power supply to the reactor coolant pumps. If the l:'eactor is at power at the time or the accident and a failure of the network bus transfel:'

occurs, the immediate effect is a more rapid increase in the coolant temperature compared to the increased coolant temperature as a result or the turbine trip by itself. The analysis assumed that the following RPS signals are available to trip the reactor: <1> undervoltage or underrrequency on the reactor coolant pump power supply busses, C2> low reactor coolant loop flow, or <3> reactor coolant pump circuit breaker opening. The reactor trip on reactor coolant pump undervoltage is provided to protect against a station blackout event and is blocked below 11% RATED THERMAL POWER Ci. e. the P-7 set point>. The reactor trip on reactor coolant pump underrrequency is provided to protect against Page 7 major power grid frequency disturbances by disengaging the pumps so that the pumps' kinetic energy is available for full coastdown.

The reactor trip on low reactor coolant flow and reactor coolant pump circuit breaker opening were discussed in Item A.1 above. Changing the turbine-reactor trip setpoint from 11% to 50% of RATED THERMAL POWER will not affect the RPS signals provided for protection against a complete loss of forced reactor coolant flow. The change removes the defeat or the automatic block or reactor trip upon turbine trip from the P-7 signal and adds the defeat to the P-9 permissive signal. This is an independent trip signal* from the three signals identified above which remain unaffected.

Between 11% and 50% or RATED THERMAL POWER the three signals identified above will still be available to mitigate the consequences or a complete loss or forced reactor coolant since the reactor coolant pumps will still be operating.

Above 50% or RATED THERMAL POWER the three signals will function as they do currently for P-7. The installation or the P-9 setpoint in no way alters or affects the capability or the RPS to

• perform its function.

As a result, the UFSAR Section 15.3.4 accident analysis bounds the consequences or a turbine trip event .below 50% of RATED THERMAL POWER with or without a subsequent reactor trip and hence the proposed changes do not increase the probability or consequences or this previously evaluated accident.

B. Do not create the possibility for a new or different kind or accident than any previously evaluated.

1. Inherent within the proposed change is the capability or the Turbine Control System to handle a steam dump to the condensor at or below 50% or RATED THERMAL POWER. UPSAR Section 10.4.4.1 discusses the design or the steam dump control system and discusses the consequences should the system fail to operate both above and below 50% or RATED THERMAL POWER. In the event or a loss or load or turbine trip at or below 50% RATED THERMAL POWER, and should the steam dump valves rail <or the condenser not be available as a heat sink>, the steam generator pressure and primary system temperature will rapidly increase.

The steam generator safety valves are sized to remove the steam flow at 105 percent or steam flow at rated power, well within the 110 percent of steam system design pressure.

With the turbine condenser not available, steam will be dumped to the atmosphere and main feedwater flow wi11 be lost. In this situation the reedwater flow will be maintained by the Auxiliary Feedwater System thereby insuring adequate residual and decay heat removal capability.

ffhen the steam generator safety valves are reseated, the pressurizer power operated relier valves <PORVs> will operate to remove residual heat and control Page 8 I

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steam pressure.

The pressurizer PORVs are sized to relieve sufficient steam to* maintain the reactor coolant system pressure within 110 percent or the reactor coolant system design pressure.

In the event or a loss or load or turbine trip above 50% or RATED THERMAL POWER or should it be necessary to close the Main Steam Stop Valves <MSSVs> under full load, safety relief valve capacity equal to 100% of full load flow is provided on the piping just upstream or the MSSVs. The capacity is provided by five self-actuated safety valves on each main steam line, with setpoints ranging from 1070 to 1125 psig, which vent via umbrella vents to* atmosphere through the roof of the penetration area. Additionally, a power operated steam relief valve is provided on each main steam line upstream of the MSSVs, total capacity for all four valves is 10% of full load. These valves have remotely variable pressure setpoints and can be used to bleed orr reactor decay heat. As discussed above, the operation of the steam dump system is not the only means to control secondary system pressure following a load loss or turbine trip. Therefore, credit can be taken for the function of this system, and hence, it can be concluded that a reactor trip below SOS or RATED THERMAL POWER is not required following a loss of load or turbine trip from a pressure standpoint.

Additionally, increasing the turbine-reactor trip setpoint from 11% to 50% of RATED THERMAL POWER does not create the possibility or a new or different accident in terms of the capability or the secondary system to handle full or partial load pressures in the event the turbine control system fails to automatically dump steam to the condenser.

2. Westinghouse has completed a study or the potential for increased pressurizer PORV opening from a turbine trip without a reactor trip at 50% or RATED THERMAL POWER and has concluded that

trip below 50% of RATED THERMAL POWER will not result in opening the pressurizer power operated relief valves, and <ii> that even considering the scenerio along with degraded control system performance

<i.e. steam dump system, pressurizer spray system or rod control system failure>, the pressurizer power operated relier valves will not open. The results or this study are presented in further detail in Attachment

2. 3. As discussed in Paragraph II above, the use or the P-9 setpoint for SGS Unit No. 1 is restricted until receipt and installation or hardware required to actually change the setpoint <currently PSE&G has only enough material onsite to install the modification for SGS Unit Ho. 1 >. As a result, PSE&G is proposing the use or the P-8 setpoint <i.e. at or below 36% or RATED THERMAL POWER> in the interim between NRC issuance or this amendment Page 9 request and the subsequent receipt and installation or the P-9 hardware.

As discussed within this submittal, the proposed change does not affect the P-7 setpoint in any manner other than to remove the turbine-reactor trip permissive.

Similarly, the addition of this permissive to the P-8 setpoint does not affect any other functions the P-8 setpoint currently performs.

Additionally, the P-8 setpoint (less than or equal to 36% of RATED THERMAL POWER> is less than the P-9 setpoint <less than or equal to 50% or RATED THERMAL POWER> and hence, the accident analyses which bound the proposed change for P-9 also bound the proposed temporary change for P-8. Therefore, the use of a P-8 setpoint prior to the P-9 setpoint on SGS Unit No. 1 will not create the potential for any new or different kind or accident than previously evaluated.

C. Do not significantly reduce the margin of safety for any Technical Specification.

The proposed change does increase the turbine-reactor trip to SOS from 111 of RATED THERMAL POWER; however, this increase is not a significant change in the Technical Specification margins or safety. This conclusion can be reached because or the inherent design feature or SGS, namely that the turbine control system is already designed such that a 50% steam dump is within the operating limits or the station. Hence, the design or the plant is not changing, only the current Technical Specification governing turbine-reactor trips. In addition, the proposed change conforms to Example 6 or 48FR14870 in that the change is clearly within all acceptable criteria with respect to the system. From the discussions provided above, PSE&G has concluded that the proposed change to the Technical Specifications does not involve a significant hazards consideration.

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.. ; *' ... * ... ""' ATTACHMENT 2 . . ' A STUDY OF THE POTENTIAL FOR INCREASED PRESSURIZER PORV OPENING RESULTING FROM TURBINE TRIP WITHOUT REACTOR TRIP BELOW 50% POWER TRANSIENT (P-9 SETPOINT STUDY) I. INTRODUCTION The Salem Units are designed with SOS load rejection As a result of this capability, the Westinghouse design criterion is that load rejections up to SOS should not require a reactor trip if all other control 1yst1ms function.properly.

Therefore, Westinghouse has proposed to implement an interlock system that would eliminate direct reactor trips on turbine trips* below SOS power, thereby decr1111ng unn&e1ssary challenges to the reactor protection system and increasing plant availability.

The NRC has expressed concerns regarding the potential increase in probability of a stuck-open pressurizer PORV following the implementation of deletion of reactor trip on turbine trip below SOS power. The NRC position is addressed in NUREG-0737, Item II-K.3.10.

The information pr1s1nt1 a best estimate ;na1yt1ca1 study to show that no additional pressurizer PORV challenges are expected due to implementation of an interlock

systa, would eliminate.

direct reactor trips on turbine trips below SOS power for Salem Untts. II. SYSTEM TRANSIENT ANALYSIS II.1 Description of the Analysis:

A best estimate analytical study was performed to determine th* transient plant response to a turbine trip without 1 reactor trip from SOS power. The analysis was performed using the LOFTRAN computer code(l) model of the Salem Units. This computer model simulates the overall thermal/hydraulic/nuclear response of the NSSS as well as the various eentrol and protection systems. Since the object of this study was primarily to determine the peak in pressurizer pr1ssur1 following the initiation of th1 transient, assumptions (1) LOFTRAN Coda Description, WCAP-7878, Rev. 0 -Rev. 3. *

. . .... *

• were made that would contribute to a conservatively high prediction of pressurizer pressure.

These assumptions were th* following.

1. Beginnfng-of-Lif* (SOL) reactivity parameters were used since this gives the *inimum moderator f1tdback, and consequently, th* minimum decr1as1 in nuclear power as a result of the initial increase in primary coolant temperature during the transient.

2. Transients were initiated from 521 power (21 calorimetric error .in advers9 direction) 1inc1 SOI power is the maximum proposed value for the P-9 permissive setpoint that would permit 1 turbine trip without actuating a direct reactor trip. Tr*nsients initiated from a lower power level would be less with roap&et to predicting th* peak . 16 pressurizer pressure.

This is true since the peak in pressurizer

-

ts directly relatid to the amount of energy that must be -. storld_tn the primary syst1m during mismatch between-cor1pow1r production and s1condary 1yst1m load. A turbine trip from 1

• initial power level simply results in a smaller power mismatch and this results in a smaller peak in pressurizer presJure.

In th1 limit, the initial power level of the transient would be reduced to lOI power which ts currently the power level at which a turbine _trip without 1 reactor trip is permitted.

3. The pressurizer lllOClel 1n LOFTRAN is conservative with respect_ to over predtcttng peaks in pressurizer pr1ssur1.

This ts because the pressurizer pressure calculation mcd&l 1n LOFTRAN ts tsentropic

• -. Comparison studt11<2> have shown that.'such models (ts1ntropfc model) ov1rpr1dtct pressurizer pressure during pr1ssur1-tncr1as1 transients.

• (2) Baron, R. c., *Digital Model Simulation of a Nuclear Pressurizer,*

Nuclear . Science and Engineering 52, 283-291 (1973). -*

I . *-*. J lI.2 Analysi_s Results: The expected system response*

to a turbine trip without a reactor trip from 50% power is shown in figures 1 through. 6 *. For nonaal plant operation with all normal systems assumed operational, the pressurizer pressuri does not !!!Eh the point of pressurizer PORV activation (PORV s1tpoint for the Salam Un1ts is 2350 psia)** Results also* indicated th1t st;am PORV does not open dYring this transi*nt.

Note that in figures 1 through 6, transient initiates after 10 seconds of steady state. III. FAILURE MODES ANALYSIS ... For normal plant operation w1th an normal control 1yst11111 aslUlllld operational, this transient does result in*opening the pressurizer PORVs. However, the NRC has expressed 1 concern that the ia.,l1111nt1tion of a turbine trip without a reactor trip be-low 50% power permissive should not result in increased challenges to the pr1sstiriz1r PORVs even tn* the event of *degraded*

control system performance

• . A sensitivity study was th1r1for1 performed in which certain failures were assumed to occur 1n the control syst1111s that influ1nc1 the course of this tr1nii1nt in order to d1t1rmin1 their 1ff1ct on the potential for pr1ssuriz1r PORV challenges.

There are thr .. main control systems that act during this transient:

the steam dump system, the pressurizer spray_ syst111 and the rod control system. The steam dump system consists of twelve valves which are arranged into four banks, three valves paF bank. A single cr1dibl1 failure was assumed to be the failure of a bank of steam dump valves to open on demand following the turbine trip. In the pressurizer spray system, the types of failure assumed were either a reduction 1n* spray flow c1pacity (du*, for 11ampla, to a sticking :pray control valve) or-1 complete failure to receive any spray flow (due, for 111mpl1, to a failure of the control actuating signal). The failure that was assumed in the rod control system was the failure of the power mismatch channel. The purpose of the power mismatch channel signal is.to provide a

• t *.:. .... _ ...... *-***-fast fHd-forward signal to the rod control 1yst1m during a rapid change fo . . . turbine load. If this signal 11 not pr1s8nt, then the rods controlled only by th* Tavg *rror signal which hu 1 much slowlir response and thus 1t takes longer time to begin driving th* r0d1 .into the core foll.owing the turbine \.rip. III.1 Failure Node Analysis Results: . The r1sult1 of the failure 11Dd1 11nsitivity atucly showed that for normal *system operation and for any single failure that was considered, both pressurizer and steam g1n1rator PORVs did not open** In fact, tt takes a . combination of multiple control syst111 failures to result 1n or ... the steam PORVi.i1i)ening during the transient.

-IV. CONCW..USIONS Based on the best estimate analysis 'results following concJusiona are made: 1. For normal plant operation with all normal control systems assumed operational, the iq>lementation of a syst1m (P-9 permissive) that permits"*

• turbine trip without actuating a direct reactor trip below SOS power w111 not result in opening the pr1ssuriz1r power-operated relief valves. 2. For any single failure in the control systllll that w1s ccnsid*rad in the 1nalyst1 1 the 1q>111n1nt1tion of a syst1111 permissive) that permits 1 turbine trip without a direct reactor trip below SOS power will not result in opening the pressurizer pow1r-operat1d r1li1f valves. It was found th1t 9 it t1k;s i ccmbinat1cn cf multiple control system failures to result in pressurizer power-operated relief valves opening during th1 transient.