JPN-88-025, Proposed Tech Specs,Changing Reactor Vessel Water Level Instrumentation Sys to Reduce Instrument Errors Resulting from post-accident Containment Heat Up & to Increase Redundancy & Reliability of Control Room Indication

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Proposed Tech Specs,Changing Reactor Vessel Water Level Instrumentation Sys to Reduce Instrument Errors Resulting from post-accident Containment Heat Up & to Increase Redundancy & Reliability of Control Room Indication
ML20154Q869
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 05/27/1988
From:
POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20154Q843 List:
References
JPN-88-025, JPN-88-25, NUDOCS 8806070151
Download: ML20154Q869 (20)
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Text

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ATTACHMENT I TO JPN-88-025 PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING REACTOR VESSEL WATER LEVEL INSTRUMENTATION (JPTS-83-014) l I

i New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 l DPR-59 l

l l

8806070151 880527 PDR ADOCK 05000333 P DCD _

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i .

JAFEPP-e 2.1 (cont'd) 4

2. Reactor Water Low Level Scram Trio Settinq l Reactor low water level scram setting shall be 2L177 in. above the top of the active fuel (TAF) l at normal operating conditions.

4

3. Turbine Stop Valve Closure Scram Trio Settina J

j Turbine stop valve scram shall be gul0 percent valve closure from full open when above 217 psig turbine first stage pressure.

1

4. Turbine Cn_ntrol Valve Fast-Closure Scram Trio
  • Settinq i

i Turbine control valve fast closure scram control oil pressure shall be set at 500 < P(850 psig.

i l

S. Main Stemm.Line Isolation Valve Closure Scram Trip Setting Main steam line isolation valve closure scram l shall ba (-10 percent valve closure from full open.

4

6. Main Steam Line Isolation Valve Closure on Low Pressurn When in the run mode main steam line low. pressure initiation of main steam line isolation valve closurs shall be jt825 psig.

Amendment No. 34, y7, 3g 11

'JAFNPP 2.1 BASES (Cont'd)

c. APRM Flux Scram Trio Setting (Run Mode) (cont'd) d. APRM Rod Block Trio Settino rated power. This reduced flow referenced trip Reactor power level may be varied by moving setpoint will result in an earlier scram during control rods or by varying the recirculation.

slow thermal transients, such as the loss of flow rate. The APRM system provides a control 80*F feedwater heating event, than would result rod block to prevent rod withdrawal beyond a with the 120% fixed high neutron flux scram given point at constant recirculation flow rate, trip. The lower flow referenced scram setpoint and thus provides au added level of protection therefore decreases the severity ( Z1 CPR) of a before APRM Scram. This rod block trip setting, slow thermal transient and allows lower Opera- which is automatically varied with recirculation '

ting Limits if such a transient is the limiting loop flow rate, prevents an increase in the abnormal operational transient during a certain reactor power level to excessive values due to exposure interval in the cycle. control withdrawal. The flow variable trip setting parallels that of the APRM Scram and The APRM fixed high neutron flux signal does not provides margin to scram, assuming a steady-

  • incorporate the time constant, but responds state operation at the trip setting, over the directly to instantaneous neutron flux. This entire recirculation flow range. The actual ,

scram setpoint scrams the reactor during fast power distribution in the core is established by l power increase transients if credit is not taken specified control rod sequences and is moni-for a direct (position) scram, and also serves tored continuously by the in-core LPRM system.

to scram the reactor if credit is not taken for As with the APRM scram trip setting, the APRM r the flow referenced scram. rod block trip setting is adjusted downward if l the maximum fraction of limiting power density The scram trip setting must be adjusted to exceeds the fraction of rated power, thus pre-ensure that the LHGR transient peak is not serving the APRM rod block margin. As with the increased for any combination of maximum scram setting, this may be accomplished by fraction of limiting power density (MFLPD) and adjusting the APRM gain.

reactor core thermal power. The scram setting is adjusted in accordance with the formula in 2. Reactor Water Low Level Scram Trio Settino  !

Specification 2.1.A.1.c, when the MFLPD is l greater than the fraction of rated power (FRP). The reactor low water level scram is ' set .at a This adjustment may be accomplished by either point which will assure that the water level (1) reducing the APRM scram and rod block set- used in the Bases for the Safety Limit is tings or (2) adjusting the indicated APRM signal maintained. The scram setpoint is based on 2

to reflect the high peaking condition. normal operating temperature and pressure

, conditions because the level instrumentation is Analyses of the limiting transients show that no density compensated.

l scram adjustment is required to assure that the i

MCPR will be greater than the Safety Limit when l the transient is initiated from the MCPR oper- ,

ating limits provided in Specification 3.1.B. ,

Amendment No. 36 i 18

a l

JAFNPP TABLE 3.1-1 (cont'd) .

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REOUIREMEriT Minimum No. Modes in Which Total of Operable Trip Level Function Must be Number of Instrument Trip Function Setting l Operable Instrument Action Channels Channels (1)-

per Trip Refuel Startup Run' Provided '

System (1) (6) by Design for Both Trip Systems 2 APRM Downscale 1 2.5 indicated on X 6 Instrument A or B scale (9) Channels 2 High Reactor 11045 psig X(8) X X 4 Instrument A Pressure Channels 2 High Drywell 1 2.7 psig X(7) X(7) X 4 Instrument A Pressure Channels 2 Reactor Low Water 2177 in. above TAF X X X 4 Instrument A Level Channels 3 High Water Level g_34.5 gallons per X(2) X X 8 Instrument A in Scram Discharge Instrument Volume Channels Volume 2 Main Steam line g_3x normal full X X X 4 Instrument- A High Radiation power background (16) Channels 4 Main Steam Line f_10% valve X(3)(5) X(3)(5) X(5) 8 Instrument A Isolation Valve closure Channels Closure Amendment No. MI, 4d , 67, 76, 97, 96 41a

t JAFNPP 3.2 BASES In addition to reactor protection instrumentation has a' direct bearing on safety, are chosen at a level which initiates a reactor scram, protective instru- away from the normal operating range to prevent inad-montation has been provided which initiates action to vertent actuation of the safety system involved and mitigate the consequences of accidents which are exposure to abnormal situations.

beyond the operator's ability to control, or termi- ^ ~

nates operator errors before they result in serious Actuation of primary containmer.t valves is initiated consequences. This set of specifications provides by protective instrumentation shown in Table 3.2-1 the limiting conditions of operation for the primary which senses the conditions for which isolation is system isolation function, initiation of the Core required. Such instrumentation must be available Cooling Systems, Control Rod Block and Standby Gas whenever primary containment integrity is required.

Treatment Systems. The objectives of the specifica-tions are to assure the effectiveness of the protec- The instrumentation which initiates primary system tive instrumentation when required, even during isolation is connected in a dual bus arrangement.

periods when portions of such systems are out of service for maintenance, and to prescribe the trip The low water level instrumentation, set to trip at settings required to assure adequate performance. 177 in. above the top of the active fuel, closes all When necessary, one channel may be made inoperable isolation valves except those in Group 1. Details of for brief intervals to conduct required functional valve grouping and required closing times are given tests and calibrations. in Specification 3.7. For valves which isolate at this level, this trip setting is adequate to prevent Some of the settings on the instrumentation that uncovering the core in the case of a break in the initiate or control core and containment cooling have largest line assuming a 60 sec. valve closing time.

tolerances explicitly stated where the high and low Required closing times are less than this.

values are both critical and may have a substantial effect on safety. The set points of other instrumen- The low-low reactor water level instrumentation is tation, where only the high or low end of the setting set to trip when reactor water level is 126.5 in, above the top of active fuel. This trip Amendment No. j#f 55

r_- . _ _ . . - _ _ - - _ _ _ _ .- __ _ -- .__ _ _

JAFNPP TABLE 3.2-1 INSTRUMENTATION THAT INITIATES PRIMARY CONTAINMENT ISOLATION Minimum Number of' Total' Number of Instrument Operable Instrument Channels Channels Provided by Design Action per Trio' System (1) Instrument Trio Level Settina for Both Trio Systems (2) 2 (6) Reactor Low Water 2*177 in, above TAF 4 Inst. Channels A Level i-' _

u 1 Reactor High Pressure f_'75 psig 2 Inst. Channels D (Shutdown Cooling Isolation) 2 Reactor Low-Low-Low jll8 in. above TAF 4 Inst. Channels A Water Level 2 (6) High Drywell Pressure f_2.7 psig' 4 Inst. Channels .A

, 2 High Radiation Main j(3'x Normal' Rated

_ 4 Inst. Channels. B l Steam Line Tunnel Full Power Background (9) 1 4

2 Low Pressure Main j[825psig(7) 4 Inst. Channels B i Steam Line j 2 High Flow Main Steam jbl40%ofRatedSteam 4 Inst. Channels B

Line Flow i

2 Main Steam Line Leak 3L40*F above max 4 Inst. Channels B j Detection High ambient Temperature 3 Reactor Cleanup Sys- j[40*F above max 6 Inst. Channels C tem Equipment Area ambient

High Temperature

, 2 Low Condenser Vacumm 2_8" Hg. Vac (8) 4 Inst. Channels B Closes MSIV's d

Amendment No. 15, M, 46, 6Y, 96, Ip3 64

_t

JAFNPP TABLE 3.2-2 INSTRUMENTATION THAT INITIATES OR CONTRQLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum No. Total of Operable Number of Instru-Instrument ment Channels Pro-Item Channels Per vided by Design for No. Trio System (1) Trio Function Trio Level Settino Both Trio Systems Remarks 1 2 Reactor Low-Low JL126.5 in. above TAF 4 HPCI & RCIC Initiates HPCI, Water Level Inst. Channe73 RCIC & SGTS.

2 2 Reactor Low-Low- jk18 in. above TAF

_ 4 Core Spray & RHR Initiates Core Spray, Low Water Level Instrument Channels LPCI, and Emergency Diesel Generators.

4 ADS Instrument Initiates ADS in conjunc-Channels tion with confirmatory low level, 120 second time delay and LPCI or Core Spray pump discharge pressure interlock if not inhibited by ADS override switches.

3 2 , Reactor High Water JE222.5 in, above TAF 2 Inst. Channels Trips HPCI Turbine and Level closes RCIC steam line isolation valve.

4 1 Reactor Low Level 2_O in. above TAF 2 Inst. Channels Prevents inadvertent (inside shroud) operation of contain-ment spray during accident condition.

Amendment No. 14, % 67, S4 66

JAFNPP TABLE 3.2-2 (cont'd)

INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum No. Total of Operable Number of Instru-Instrument ment Channels Pro-Item Channels Per vided by Design for Do. Trip System (1) Trip Function Trip Level Setting Both Trip Systems Remarks 3 2 Containment High 1<p42.7 psig 4 Inst. Channels Prevents inadvertent Pressure operation of containn.ent spray during accident condition.

6 1 Confirmatory Low [177in. above TAF 2 Inst. Channels ADS Permissive in Level conjunction with Reactor Low-Low-Low Water Level.

7 2 High Drywell 2.7 psig HPCI Inst. Channels Initiates Core Spray Pressure LPCI, HPCI and SCTS.

8 2 Reactor Low Pres- 1450psig 4 Inst. Channels Permissive for opening sure Core Spray and LPCI Admission valves.

ATsndment No. 14, A8,f/f, 74 67

q l

TABLE 3.2-6 ,

SURVEILLANCE INSTRUMENTATION Minimum No.

of Operable No. of Channels Instrument Type Indication Provided Channels Instrument and Rance by Desian Action Narrow Range Reactor Level Indicator (13) (2)

(Note 3) 164.5 to 224.5 in. above TAF 2

Narrow Range Recorder 3' Reactor Level 164.5 to 224.5 in. above TAF (Note 4)

Wide Range 1 Reactor Level Indicator (2)

(Note 14) 14.5 to 224.5 in. above TAF 2

Wide Range Indicator-Recorder Reactor Level 14.5 to 224.5 in, above TAF (Note 15) 1 Fuel Zone Indicator (2)

Reactor Level 150 in. below to 200 in.

(Note 16) above TAF 2 Fuel Zone Indicator-Recorder Reactor-Level 150 in. below to 200 in.

(Note 17) above TAF Reactor Pressure Indicator (Note 5) 0-1200 psig 2 Reactor Pressure Recorder 5 (1) (2)

(Note 6) 0-1200 psig Amendment No. 33, Sir, 57, Ff, p9 76

M TABLE 3.2-6 (Cont'd) .

SURVEILLANCE INSTRUMENTATION Minimum No.

of Operable No. of Channels Instrument Type Indication Provided Channels Instrument and Ranoe by Desian Action Drywell Pressure (Narrow Range)

(Narrow Range) Indicator Recorder-10 - 19 psia 1 Drywell Pressure (Wide Range) 2 (2)

Indicator Recorder 0 - 100 psia Drywell Temperature Indicator 50 - 250*F 2 4 (1) (2)

Drywell Temperature Recorder 50 - 350*F Suppression Chamber Indicator Temperature 50 - 250*F 2 4 (1) (2)

Suppression Chamber Recorder Temperature 50 - 350*F Amendment No. 51, 6Y, f4 76a 9

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y NOTES FOR TABLE 3.2-6 (CONTINUED) ,

14. One (1) indicator from reactor wide range' level instrument channel A.
15. One (1) indicator-recorder from reactor wide range level instrument channel B.

1 3 16. One (1) indicator from reactor fuel zone level instrument channel A.

17. One (1) indicator-recorder from reactor fuel zone level instrument channel 3._

1 1

l I

l l

i!

1 Amendment No.

76d

.)

1 JAFNPP TABLE 3.2-7 INSTRUMENTATION THAT INITIATES RECIRCULATION PUMP TRIP Minimum Number of Total Number of Instrument Opsrsble Instrument Channels Provided by Chi.nnels per trip Design for Both System (1) Instrument Trip Level Setting Channels Action 1 Reactor High Pressure [1120psig 4 (2) 1 Reactor Low-Low 1126.5 in. above TAF 4 (2)

Water Level rotes for Table 3.2-7

1. Whenever the reactor is in the run mode, there shall be one operable trip system for each parameter for each operating recirculation punP. From and after the time it is found that this cannot be met, the indicated action shall be taken.
2. Reduce power and place the Mode Selector Switch in a Mode other than the Run Mode within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Amendment No. Ff, p6 77

1 ATTACHMENT 11 TO JPN-88-025 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING REACTOR VESSEL WATER LEVEL INSTRUMENTATION (JPTS-83-014) f New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59

Attachment 11 to JPN-88-025 SAFETY EVALUATION l Page 1 of 7 I. DESCRIPTION OF TIIE PROPOSED CIIANGES The proposed changes to the James A._ FitzPatrick Technical Specifications revise Specification and Bases 2.1.A.2, Bases 3.2, Tables 3.1-1, 3.2-1, 3.2-2, 3.2-6, and 3.2-7 on pages 11,18, 41a, 55, 64, 66, 67, 76, 76a, and 77, and adds a new page 76d. Changes, identified by a capital letter in brackets, reflect modifications to the Reactor Vessel Water Level Instrumentation system and elimination of obsolete level setpoints. The changes are as follows:

[A] Page 11, Specification 2.1.A.2 Reactor Water Low Level Scram Trip Setting (LLI)

1) Delete " (LLI) " from title.
2) Delete " (+12.5 in. indicated level) "

[B] Page 18, BASES 2.1.A.2 Reactor Water Low Level Scram Trip Setting (LLI)

1) Delete " (LL1) " from title.

[C] Page 41a, TABLE 3.1-1 (cont'd)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Trip level Setting for Reactor Low Water Level:

1) Delete " > 12.5 in. indicated level"
2) Remove parentheses

.. Replace "the top of active fuel" with "TAF*

[D] Page 55, BASES 3.2 PROTECTIVE INSTRUMENTATION Delete " (-38 in, on the instrument) " from the sixth parapraph.

[E] Page 64, TABLE 3.2-1 INSTRUMENTATION TIIAT INITIATES PRIMARY CONTAINMENT ISOLATION Trip Level Setting for Reactor Low Water Level:

1) Delete " > 12.5 in, indicated level"
2) Remove parentheses
3) Replace "the top of active fuel" with "TAF" Trip Level Setting for Reactor Low-Low-Low Water Level:
4) Replace "the top of active fuel" with "TAF"

[F] Pages 66 and 67, TABLE 3.2-2 INSTRUMENTATION TilAT INITIATES OR CONTROLS Tile CORE AND CONTAINMENT COOLING SYSTEMS Trip Level Setting for Reactor Low-Low Water Level

1) Delete " > -38 in. indicated level"
2) Remove parentheses
3) Replace "the top of active fuel" with "TAF"

Attachment Il to JPN-88-025 SAFETY EVALUATION Page 2 of 7 Trip Level Setting for Reactor Low-Low-Low Water Level

4) Delete "1-146.5 in. indicated level"
5) Remove parentheses
6) Replace "the top of active fuel" with "TAF" Trip Level Setting for Reactor High Water Level
7) Delete "1+58 in. indicated level"
8) Remove parentheses
9) Replace "the top of active fuel" with "TAF" Remarks for Reactor liigh Water Level
10) Replace "Trips llPCI and RCIC Turbines" with "Trips 11PCI Turbine and closes RCIC steam line isolation valve."

Trip Level Setting for Reactor Low Water Level (inside shroud)

11) Delete "1+352 in, above vessel zero"
12) Remove parentheses
13) Replace "the top of active fuel" with "TAF" Trip Level Setting for Confirmatory Low Level
14) Delete "1 12.5 in. indicated level"
15) Remove parentheses
16) Replace "the top of active fuel" with "TAF" (G) Pages 76 and 76a, TABLE 3.2-6 SURVEILLANCE INSTRUMENTATION Entry for Reactor Level (first two occurrences)
1) Insert "Narrow Range" above "Reactor Level"
2) Delete "O - +60"
3) Remove parentheses
4) Replace "the top of active fuel" with "TAF"
5) No. of Channels Provided by Design is changed from *5" to "3" Entry for Reactor Level (third occurrence)
6) Insert "Wide Range" above "Reactor Level" i
7) Insert " (Note 14) " under "Reactor level"
8) Delete "- 150 - +60" l 9) Femove parentheses
10) Replace "the top of active fuel" with "TAF"
11) Add an entry for a new instrument to read:

! Wide Range Indicator-Recorder Reactor Level 14.5 to 224.5 in, above TAF (Note 15) l l

Attachment 11 to JPN-88-025 SAFETY EVALUATION Page 3 of 7

12) Add new table entries for the fuel zone level instruments as follows:

Minimum No. of Operable Instrument Channels: 1 Instrument: Fuel Zone Reactor Level (dote 16)

Fuel Zone Reactor Level (Note 17)

Type Indication and Range: Indicator 150 in, below to 200 in above TAF Indicator-Recorder 150 in. below to 200 in. above TAF No. of Channels Provided by Design: 2 Action- (2)

Table Entry for Drywell Pressure

12) This entry remains unchanged, but is relocated onto page 76a.

[11] Page 76d (new). NOTES FOR TABLE 3.2-6 (CONTINUED)

Four notes (14 through 17) are added to describe the instrument channel sources for the instrumentation added are changed in changes F.6,10, and 11 above.

[J] Page 77, TABLE 3.2-7 INSTRUMENTATION TilAT INITIATES RECIRCULATION PUMP TRIP Trip Level Setting for Reactor Low-Low Water Level

1) Delete "> -38 in. indicated level"
2) Remove parentheses
3) Replace "the top of active fuel" with "TAF" II. PURPOSE OF TIIE PROPOSED CIIANGES The proposed changes to the Technical Specifications fall into three categories:

A) Changes to reflect the installation of new instrumentation; B) Changes to reflect the elimination of obsolete instrument setpoints and other miscellaneous items; and C) One change to reflect a plant modification performed to meet the requirements of NUREG-0737 Item II.K.3.13.

CATEGORY A in accordance with the requirements of Regulatory Guide 1.97 and Generic Letter 84-23, the reactor water level instrumentation system is being modified. New instrumentation is being installed and the ranges of several instruments are increased. Additionally, the reference legs for the wide range instruments are to be relocated outside of the containment. (This does not regt' ire any changes to the Technical Specifications.) The following changes fall into this category: G.4, G.ll-13, and 11.

Attachment 11 to JPN-88-025 SAFETY EVALUATION Page 4 of 7 CATEGORY B The FitzPatrick Technical Specifications currently use two sets of reactor vessel instrumentation setpoints. The first set is the original setpoints. These setpoints are referenced to either the vessel bottom or to the bottom of the steam separator skirt. The second set, added in Reference 4 references all reactor water levels to the top of active fuel. The proposed change removes the original setpoints. The following changes fall into this category: A.2, C, D, E, F.1-9, F.11-16, G.1-4, G.8-10, and J.

Several miscellaneous changes are also made to the Technical Specifications. Changes A.1 and B.1 remove a label (LL1) that is not used at FitzPatrick. Change G.5 corrects an error contained in the Specifications. The 9-05 panel in the FitzPatrick Control Room contains three narrow narrow range reactor water level indicators. One of two of these signals can be selected to drive a pen on a chart recorder. Since there are only three sensors which drive these instruments, the number of instrument channels provided by design is three. This configuration is described in existing Notes 3 and 4 to the table. The error was introduced into the Technical Specifications in an application for amendment to operating license (Reference 10) and issued by the NRC in Amendment No. 48 (Reference 11).

The purpose of these changes is to remove the possibility of operator error due to the dual reference system or reliance on incorrect information. The proposed change improves the clarity and consistency of the Technical Specifications.

CATEGORY C This change (F.10) reflects a change in the RCIC logic design. Upon receipt of a reactor high water level signal, the turbine steam supply isolation valve 13-MOV-131 will auto-close.

Previously, this signal tripped the turbine by closing the hydraulic trip valve. Resetting this trip valve required local manual action, effectively making RCIC unavailable for the remainder of the transient.

The purpose of the change is to allow the RCIC system to auto-restart upon receipt of a subsequent reactor low water level signal. The intent of the existing specification is to protect the RCIC turbine from water admission. This can be accomplished by closure of any of the steam line valves. Since MOV-131 will reopen upon a RCIC initiation signal, it was selected to also close on high water level to provide the turbine protection function.

This modification was installed ir 1981 to meet the requirements of NUREG-0737 Item II.K 3.13 (Reference 7). The Authority provided a detailed description of the modification to the NRC in Reference 8. The NRC reviewed the modification and found it to be acceptable (Reference 9).

III. IMPACT OF TiiE PROPOSED CIIANGES CATEGORY A The reactor vessel water level instrumentation system is being modified to comply with NRC Generic Letter 84-23 and Regulatory Guide 1.97 Revision 2. No changes are made to the actual reactor water levels at which safety actuations occur. A detailed description of the modification is contsined in Reference 5. The modification includes the following:

Attachment 11 to JPN-88-025 -

SAFETY EVALUATION Page 5 of 7

1) Replacing the existing wide-range level instrument's temperature compensated reference leg system with a cold leg system located outside the containment drywell. This will reduce level indication and ECCS initiation errors caused by high drywell temperatures during certain postulated accident conditions. High drywell temperature can be caused by the loss of drywell coolers, or by high energy line breaks (including loss of coolant accidents).

When the reactor is depressurized to the saturation pressure of the water in the sensing lines, or the drywell heats up to the saturation temperature, flashing of water in the lines will occur and some of the water in the reference legs inside the drywell will boil off.

Loss of water from the reference legs results in an indicated water level that is higher than actual.

2) One of the wide-range level indicators in the control room will be replaced with a level indicator-recorder. This provides a permanent record of the reactor water level as sensed by the wide-range instruments.
3) Two of the wide-range level instrument loops will be upgraded to conform to redundant Class IE requirements. This increases the availabilty of control room indication under certain loss of electrical bus conditions.
4) The fuel zone instruments will be recalibrated to extend their range to the bottom of active fuel (BAF). Technical Specifications are also to be applied to these instruments.

This assures the ability to monitor and record the reactor water level under postulated accident conditions when the water level is significantly below TAF. Technical Specifications will assure the operability and availability of these instruments.

CATEGORY B NUREG-0737 Item II.K.3.27 (Reference 3) required all boiling water reactors to change the reactor vessel water level instrumentation reference system. Amendment No. 67 (Reference 4) added a common reference point (top of active fuel) to each reactor level instrument setpoint in the Technical Specifications to fulfill this NUREG-0737 requirement.

Plant procedures, training programs, control room and local instrumentation have been or will be changed to reflect the new instrument reference point. The old water level references have been rendered obsolete since the issuance of Amendment No. 67. Removing them from the Technical Specifications improves consistency and reduces the probability of misinterpretation.

No hardware or procedural changes (except for changes in nomenclature) are required to implement the proposed changes. Nor are any changes made to the actual reactor water levels at which safety actuations occur.

CATEGORY C NUREG-0737 Item II.K.3.13 (Reference 7) required all boiling water reactors to modify the RCIC system such that the system will restart on subsequent low water level after it has been terminated by a high water level signal. The FitzPatrick RCIC system was modified in 1981 to meet this requirement. The modification is described in Reference 8 and approved by the NRC in Reference 9.

The proposed change to the Technical Specifications reflects this modification to the RCIC logic and trip valve design. Previously, a high reactor water level signal tripped the RCIC turbine by closing the hydraulic trip valve. Resetting this trip valve required local manual action, effectively making RCIC unavailable for the remainder of the transient. The purpose of the change is to allow RCIC to automatically reinitiate on low reactor water level following a high water level trip. This change contributes to improved system reliability during transients or postulated accident conditions.

~

Attachment Il to JPN-88-025 SAFETY EVALUATION Page 6 of 7 The intent of the high water level trip is to isolate the RCIC steam supply line to prevent water from being introduced into the turbine. This turbine protection function can be performed by closing any valve in the steam line. Therefore, the proposed change does not change the intent of the specification.

If water was to enter the steam supply line prior to the reinitiation of RCIC, the steam trap located just ahead of the isolation valve should drain the line. RCIC was designed to withstand introduction of water into the turbine. If the line was not completely drained, no damage should result to the RCIC system.

No credit is taken for RCIC operation in the transient and c. Mat analyses contained in the FitzPatrick FSAR. Therefore, modifications to the RCIC logic cannot impact any margin of safety or analysis as contained in the FSAR.

IV. EVALUATION OF SIGNIFICANT IIAZARDS CONSIDERATION Operation of the FitzPatrick Plant in accordance with the proposed Amendment would not involve a significant hazards consideration as stated in 10 CFR 50.92 since it would not:

1. involve a significant increase in the probability or consequences of an accident previously evaluated. The Category A changes improve the reliability and availabilty of reactor water level signals which are used to initiate ECCS system and provide control room indication. The modification is not an accident initiator. The Category B change is purely administrative in nature and does not involve a physical change to the facility, and therefore, cannot affect any accident as analyzed in the FSAR. The Category C change improves the reliability and availability of a system which can be used under accident conditions. No credit has been taken for this system in the FSAR accident analyses. liowever, it improves the reliability of a system which could be used to mitigate the probability or consequences of accidents.
2. create the possibility of a new or different kind of accident from any accident previously evaluated. As stated above, the proposed changes increase the reliability of systems or functions which mitigate accident conditions (Category A and C) or are purely administrative in nature (Category B). No new or different types of accidents can occur as a result of improving the availability and reliability of the reactor water level instrumentation system, eliminating obsolete instrumentation set point references, or improving the availability of the RCIC system.
3. involve a significant reduction in a margin of safety. The proposed changes improve the performance of instrumentation and systems which mitigate transients and accidents.

Elimination of obsolete reactor water level instrument zeros achieves consistency in the reactor vessel instrumentation setpoints. This reduces the probability of a misinterpretation of the specifications. No reduction of any margin of safety is caused by the proposed changes.

V. IMPLEMENTATION OF TIIE PROPOSED CIIANGE Implementation of the proposed changes will not impact the ALARA or Fire Protection Programs at FitzPatrick, nor will the changes impact the environment.

Attachment 11 to JPN-88-025 SAFETY EVALUATION Page 7 of 7 VI. CONCLUSION The chanBe, as proposed, does not constitute an unreviewed safety question as defined in 10 CFR 50.59. That is, it:

a. will not change the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report;
b. will not increase the possibility of aa accident or malfunction of a different type from any previously evaluated in the Safety Analysis Report;
c. will not reduce the margin of safety as defined in the basis for any technical specification;
d. does not constitute an unreviewed safety question; and
e. involves no significant hazards consideration, as defined in 10 CFR 50.92.

VII. REFERENCES

1. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report, Table 7.3 -3.
2. James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER).
3. NRC NUREG-0737 Post TMI Requirements, Action Item II.K.3.27 "Provide Common Reference Level for Vessel Level Instrumentation."
4. Letter, P.J. Polk (NRC) to L.W. Sinclair (NYPA), dated February 26, 1982, issuing Amen:iment No. 67 to the FitzPatrick Technical Specifications.
5. Letter, J.C. Brons (NYPA) to NRC, JPN-87-033, dated June 9, 1987, concerning implementation of NRC Generic Letter 84-23.
6. Regulatory Guide 1.97, Revision 2.
7. NRC NUREG-0737 Post TMl Requirements .a II.K.3.13 "I1PCI/RCIC Initiation Level."
8. Letter, J.P. Bayne (PASNY) to D.B. Vassallo (NRC), JPN-83-036, dated April 28, 1983, concerning NUREG-0737 Item II.K.3.13 - RCIC Automatic Restart.
9. Letter, D.B. Vassallo (NRC) to J.P. Bayne (PASNY), dated June 17, 1983, concerning TMI Action Plan Item II.K.3.13, llPCI/RCIC Initiation Level.
10. Letter, P.J. Early (PASNY) to T.A. Ippolito (NRC), JPN-79-058, dated September 13, 1979, concerning Propsed Technical Specifications Changes Related to Instrumentation,
11. Letter, T.A. Ippolito (NRC) to G.T. Berry (PASNY), dated January 23, 1980, issuing Amendment No. 48 to the FitzPatrick Technical Specifications.
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