ML19332G367

Proposed Tech Spec Table 3.2-2, Instrumentation That Initiates or Controls Core & Containment Cooling Sys & Associated Bases,Changing HPCI Steam Line Flow Trip Level Setting

ML19332G367
Person / Time
Site:	FitzPatrick
Issue date:	12/11/1989
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML19332G363	List: ML19332G362 ML19332G366 ML19332G367
References
JPTS-89-037, JPTS-89-37, NUDOCS 8912210158
Download: ML19332G367 (8)

Text

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l ATTACHMENTI PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING HPCI STEAM LINE ISOLATION SIGNAL  !

l (JPTS 89-037) t l

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

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JAFNPP 32 BASES (cont'd)

High radiation monitors in the main steam line tunnel have The trip settings of approximately 300 percent of design flow l been provided to detect gross fuel failure as in the control rod for this high flow or 47F above maximum sidient for high drop accident. With the established setting of 3 times normal temperature are such that uncovenng the core is prevented background, and main steam line isolation valve closure, and fission product release is within limsts.

fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident. Reference Section 14.6.12 W OC@ h M w he'N M'me arranged the same as that for the HPCI. The trip setting of FSAR. During the Hydrogen Addition Test, the normal background Main Steam Line Radiation Level is expected t

,.W @ 300 p M a @ h W M h l-maximum ambient for temperature are based on the same increase by approximately a factor of 5 at the peak hydrogen criteria as the HPCI~

concentration as indicated in note 16, Table 3.1-1. With the hydrogen addition, the fission product release would still be The reactor water cleanup system high flow temperature well within the 10 CFR 100 guidelines in the event of a control instrumentation are arranged similar to that for the HPCI. The rod drop accident. trip settings are such that uncovering the core is prevented and Pressure instrumentation is provided to close the main steam

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isolation valves in the run mode when the main steam line The instrumentation which initiates ECCS action is arranged in pressure drops below 825 psig. The reactor pressure vessel a dual bus system. As for other vital is astrumentation arranged thermal transient due to an inadvertent opening of the turbine in this fashion, the specifk.aiion preserves the effectiveness of bypass valves when not in the run mode is less severe than the the system even during periods when marntenance or testing is loss of feedwater analyzed in Section 14.5 of the FSAR, being performed. An exception to this is when logic ftn,ctional therefore, closure of the main steam isolation valves for thermal testing is being pericanisd.

transient protection when not in the run mode is not required.

The control rod block functions are provided to prevent The HPCI high flow and temperature instrumentation are excessive control rod withdrawal so that MCPR does not de-provided to detect a break in the HPCI stea n piping. Tripping of this instrumentation results in actuation of HPCI isolation valves. Tripping logic for the high flow is a 1 out of 2 logic.

-Amendment No. 14,37,28,46,56,Jgt

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1 JAFNPP TABLE 32-2 (Cont'd)

INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minirmm No.

of Operable Totai Number of Instrument Instrument Channels item Channels Per Provided by Design No. Trip System (1) Trip Function Trip Level Setting for Both Trip Systems Remarks 29 1 RCICSteam Une/ <407 Above 2 Inst. Chc= wiels Closeisolation Valve Area Temperature max. EzTine:Ts in RCIC Subsystem 30 1 RCIC Steam Une 100>P>50 psig 2 Inst. Channels Close isolationValves Low Pressure in RCICSubsystem 31 1 HPCI Turbine Steam <160 in H2 O dp 2 Inst. Channels CloseIsolation Valves I Une High Flow in HPCI Subsystem 32 1 RCIC Turbine High <10 psig 2 Inst. Channels Closeisolation Valves Exhaust Diaphragm in RCIC Subsystem Pressure 33 1 HPCI Turbine High < 10 psig 2 Inst. Channels Close isolation Valves Exhaust Diapinown-i in HPCI Subsystem Pressure 34 1 LPCI Cross-Connect NA 1 Inst. Channels initiatesannunoation Position when valveis not closed 35 1 HPCI Steam Une 100> P>50 psig 2 Inst. "hs Close Isolation Valve Low Pressure in HPCISubsystens SS 1 HPCI Steam Une/ <40T 2 inst. Channels Close Isolation Valve

. Area Temperature above max. ambient in HPCI Subsystem

. Amendment No. 4,46,66, g4 70b

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ATTACHMENT 11 SAFETY EVALUATION FOR  !

PROPOSED TECHl4T6K[lPECIFI6ATi6R'6RINGES REGARDING i HPCI STEAM UNE ISOLATION SIGNAL ,

i (JPTS-89 037) l 5

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

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. Attachment ll SAFETY EVALUATION )

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1. DESCRIPTION OF THE PROPOSED CHANGES j The proposed changes to the James A. FitzPatrick Technical Specifications revise the trip level setting for the HPCI turbine steam line high flow in item 31 of Table 3.2 2 on page 70b and the associated Bases on page 57. This setpoint is revised from the current value of $106 in. H2 O dp to $160 in. H O 2 dp. The change to the Bases would clarify the required trip setting to be '

- approximately equal to 300 percent of rated steam flow.

11. PURPOSE OF THE PROPOSED CHANGES The purpose of the High Pressure Coolant injection (HPCI) isolation is to limit reactor coolant ,

inventory loss and the offsite radiological consequences in the event of a break in the HPCI

• piping. The analytical limit for determining if a break exists in the HPCI system piping is ste.tod in FSAR Section 7.4.3.2.7 as HPCI steam line flow of 300 percent of rated. At the FitzPatrick plant, the HPCI steam line flow rate is sensed by differential pressure elbow taps on the HPCI steam line. Recent tests han shown that the technical specification differential pressure setpoint actually corresponda it., an indicated steam line flow of approximately 200 percent of rated.

Sinco the start-up pressure transient can result in an indicated steam flow higher than 200 percent of rated, this instrumentation has generated inadvertent isolations, and needs to be amended.

The purpose of the proposed change is to reviso the trip setting for the HPCI turbine steam line high flow isolation signal and clarify the required trip setting associated with this signal. "

Inadvertent isolation has required the HPCI system to be declared inoperablo. With the technical 3pecification setpoint changed. HPCI would be fully operable. The basis for the revised setpoint is contained in Section lli beloa The FitzPatrick plant is currently operating in a 7 day Umiting Conditions for Operation (LCO) under the provisions of Specific tion 3.5.C.1.a. This LCO expires at approximately 6:00 am. on Tuesday, December 12,1089, if this change is not granted prior to this time, then Specification 3.5.C.1.b requires the plant to be in a cold condition, less than 150 psig reactor pressure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The Authority requests that this application for amendment be processed on an emergency basis under the provisions of 10 CFR 50.91(a)(5), since insufficient time exists to provide for a 30 day public comment period without requiring a plant shutdown. Justification for the emergency classification is provided in Section IV below.

If this amendment cannot be granted prior to the expiration of the 7 day LCO, the Authority requests a temporary waiver to Specification 3.5.C.1.b allowing continued full power operation until the amendment is granted.

Ill. IMPACT OF THE PROPOSED CHANGES The most significant impact of the proposed change would be to allow the FitzPatrick plant to continue full power operation. The amendment would allow the Authority to declare the HPCI system operable and exit from the LCO conttJned in Technical Specification 3.5.C.1.a.

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. Attachment 11 y

SAFETY EVALUATION Page 2 of 4 General Electric has provided guidance in Reference 3 for calculating the appropriate technical specification HPCI high steam flow instrument setpoint. This method uses measured data from the as-built HPCI system to determine the rated flow elbow tap instrument reading, and then applies several factors to calculate the elbow tip instrument reading corresponding to 300 percent rated steam flow. This calculated instrument reading yields the proper technical specification limit. Based upon this methodology, the trip setting for the FitzPatrick HPCI system should be 160 in. H2 O dp, rather that the current value of 106 in. H2 O dp.

The 300 percent flow analytical limit for defining the instrument setpoint is an historically accepted value selected to be above potential transient indicated flow rates which can occur during system startup but remains much below expected flows due to a pipe break. Because the 300 percent value is an acsigned conservative limit which is not the result of FSAR design basis calculations, minor variations are not a safety concem (Reference 3).

The Authority is also requesting that if this amendment cannot be granted prior to the expiration 1 of the LCO, then the LCO be extended until it is granted. Since the nature of the HPCI inoperability is only apparent when the reactor is at full pressure (1000 psig), the shutdown requirement (<150 psig) does not permit HPCI to be adequately tested to determine operability, in addition, the HPCI system is fully operational in a manual or auto mode should the need exist.

IV. EVALUATION OF EMERGENCY SITUATION As stated in Section 11 above, the Authority requests that this application for amendment be processed on an emergency basis under the provisions of 10 CFR 50.91(a)(5), since insufficient time exists to allow for a 30 day public comment period without requiring a plant shutdown.

l 10 CFR 50.91(a)(5) defines an emergency situation as:

I "... failure to act in a timely way would result in dorating or shutdown of a nuclear power plant, or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power lovel, ." / emphasis  ;

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The proposed Technical Specification change is required to allow the Fitzpatrick plant to continue full power operation. if the amendment is not lesued, the HPCI system would continue to be considered inoperable, requiring the FitzPatrick plant to commence shutdown at 6:00 a.m.

December 12, and be in a cold condition, at less than 150 psig reactor pressure within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> under the provisions of Specification 3.5.C.1.b.

10 CFR 50.91(a)(5) requires the Authority to address the following concem before the NRC can make a determination that an emergency condition exists:

"Whenever an emergency situation exists, a licensco requesting an amendment must explain why this emergency situation occurred and why it could not avoid this situation ..."

During the recently completed 1989 mid-cycle maintenance outage, a significant amount of maintenance, including replacement of the HPCI turbine controller was performed. Upon return to service, HPCI wcs testod in accordance with Technical Specification 4.5.C.1. HPCI faileo this

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SAFETY EVALUATION  :

Page 3 of 4 test due to an auto-isolation on high steam line flow. The cause of this isolation was attributed to .

improper venting of the Instrumentation. Since subsequent testing was successful, no additional investigation was performed. l Repeat et the same isolation signal during the December 4 test led the Authority to investigate I other possible causes of the event. The HPCI system was then put through a series of tests with additionalinstrumentation installed to isolate the root cause of the inadvertent isolations. At this time, the proximity of the normal start-up steam flow transient to the isolation setpoint was detected and confirmed. Because the installed steam line elbow dp instrumentation used to sense steam flow does not behave linearly with the actual start up st6am flow transient, the possibility presently exists that the instrumentation may generate an isolation signal during any normal HPCI start up. This apparent randomness to the generation of isolation signals made it <

very difficult to isolate the root cause of the inadvertent isolations and the necessity to use more extensive temporary test instrumentation and detailed data analysis.

V. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the FitzPatrick plant in accordance with the proposed Amendment would not involve a significant hazards corbideration 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 HPCI system is designed to mitigate the consequences of a small break Loss of-Coolant accident and as backup to the Reactor Core Isolation Coo,ing (RCIC) system for Loss of Foodwater transients by providing a source of high pressure coolant injection into the reactor. The proposed change will reduce the probability that the HPCI system would be unavailable upon demand due to inadvertent isolation.

Changing the HPCI high steam flow Instrument setting cannot initiate any previously evaluated accidents. The probability and consequences of a HPCI steam line break going undetected or unisolated is unchanged, because this brehk would result in stcam flows significantly in excess of the proposed isolation setpoint.

The 300 percent rated flow isolation slgrJ .s relied upon in the FitzPatrick high energy line break (HELB) analysis to limit the amount of steam released into the reactor building.

l Since the proposed amendment does not change the 300 percent analytical limit, it does j not result in a change to HELB environmental parameters for the reactor building or affect any of the results of the HELB analysis.

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2. create the possibility of a new or different kind of accident from any accident previously evaluated.

Changing the HPCI high steam flow instrument setting cannot initiate any type of accident.

The HPCI system is designed for standby service and only functions to mitigate accidents L

by providing a high pressure source of water to the reactor. No new failure modes are l created by the proposed change in instrument setpoint.

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. SAFETY EVALUATION Page 4 of 4

3. involve a significant reduction in a margin of safety.

The proposed change provides greater assurance that the HPCI system would be available to mitigate an accident as designed. HPCI would also be less likely to fail surveillance tests due to inadvertent isolations. This improved reliability reduces the  ;

number of unneensary tests performed on the HPCI and other ECCS systems as required when HPCI is inoperable. There is no reduction of any margin of safety.

VI. IMPLEMENTATION OF THE PROPOSED CHANGE implementation of the proposed changes will not impact the ALARA or Fire Protection Programs at FitzPatrick, nor will the changes impact the environment.

Vll. CONCLUfj!ON The change, 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 en accident or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report;

b. will not increase the possibility of an accident or malfunction of a different type from any previously evaluated in the Safety Analysis Report; P

c. will not reduce the margin of safety as dehned 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.

Vill. REFERENCES

1. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report.

2. James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER), dated November 20,1972, and Supplements.

3. General Electric Service Information Letter 475 Rev. 2, 'RCIC and HPCI High Steam Flow Analytical Umit," dated November 28,1988.

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