ML20101S564

Proposed Tech Spec Change,Correcting Error in Table 3.7-1, Process Pipeline Penetrating Primary Containment. Safety Evaluation Encl

ML20101S564
Person / Time
Site:	FitzPatrick
Issue date:	01/30/1985
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20101S530	List: ML20101S527 ML20101S543 ML20101S564
References
JPN-85-08, JPN-85-8, NUDOCS 8502050497
Download: ML20101S564 (8)

Text

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ATTACHMENT I to JPN-85-08 PROPOSED TECHNICAL SPECIFICATION CHANGES RELATED TO PROCESS PIPING PENETRATING PRIMARY CONTAINMENT (JPTS-85-003)

NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59 0

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, FASLE 3.7-1

• peDCSSS PIPEL198 PSIItfA4T!s3 PRIN4Af COerFAIIIsIEstf (sembers in parentheems are keyed to noters on totlowtag pages equal codes are LLoted codes are listed on following pages) .

Power Location Power Drywo11 valve type to Open set. to to Close Isolation closing Wormal Remarks and Line Isolated Penetration (6) ($) (6) Group Drywell (5 ) (6) Signal Time (7) Status Enceptions gesta Steen Line I-74,3,C.D AO Globe Air and 4 Inside Air and 8,C D.P,5 Note (1) Open AC, DC spring stain Steam Line E-74,5,C.D 40 Globe . Air and A Outside Air and 3,C,D,P,5 Wote (1) Open AC, DC spring anta steam LLae x-e seo Gate AC 4 Inside AC e,C.D,P,5 15 sec closed Orata Mata Steam Line E-e NO Gate DC A Outside DC 8,C,D,P.E 15 sec - Closed Deals From meector X-94, S Check - A Outstee Process Rev. flow II4 Open Feeesater From Reector 1-94, S Check - A laside Process Rev. flow IIA Opes reeemeter anactor Water X-41 40 Globe Air and & Inside Spring 3,C M4 Opea sample AC menetor deter 1 41 40 Globe Air and 4 Outside Spring 8,C WA Open Sample AC Control Rod Hy- 1-36 Check - A Inside Process Rev. flow NA) draulic Retarm I control mod asy- x-36 Check - 4 Outside reocess new. flow seal Opens on mod draulic Betura i movement and

) closed at all I other times,ssote (4) 3 control med 1-38 50 Valves 4Lr and A Outside Spring Note ( 4 8 NAl Drive Exhaust AC 1 I

I Cetrol RM 1-38 30 Valves Lar end A Outside Spring Note ( 4 ) N4)

Cetwe EsMast 4C 3

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Control Rod 1-37 SO Valves 44r and 4 Outside 3pring Note (4) N4)

Drive Inlet 4,' l

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Control aoa E-37 30 Valves Air amt & Outside Spring Note (4) MAI Deswe Inlet g N. N 1%

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9 ATTACIOtENT II to JPN-65-08 SAFETY EVALUATION RELATED TO PROCESS _EIPING PENETRATING PRIMARY CONTAINMENT (JPTS-85-003) i NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59 i

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3 I. Descrintion of the Chance This proposed Amendment revises Table 3.7-1 (" Process Pipeline Penetrating Primary Containment") on page 198 to Appendix A of '

the James A. FitzPatrick Nuclear Power Plant Operating '

License. The isolation signals for two reactor water sample valves (drywell penetration X-41) are changed from "B. C. D E.

& P" to "B & C". ,

II. Purnose of the Chance The purpose of this change is to correct an error in Table 3.7-1 that was inadvertently introduced during the initial issuance of the FitzPatrick Technical Specifications (Appendix A). During the preparation of the original issuance of the Technical Specifications, three (3) additional isolation signals ("D E, & P") were incorrectly added to the reactor water sample line (Penetration X-41) in the Table 3.7-1. This error was discovered during normal operation when the reactor water isolation valves did not isolate on all signals listed in l

Table 3.7-1.

In Reference (a), the Authority incorrectly described the error 2

e in Table 3.7-1 as a design change made during construction that had not been incorporated in the Technical Specification or Final Safety Analysis Report. However, the original plant design for the isolation of the reactor water sample line only called for isolation signals "B & Ca. This is consistant with I General Electric design of other BWR type plants. The l

incorporation of these additional isolation signals is an administrative error in both the Technical Specifications and ,

the Final Safety Analysis Report. <

III. Innact of the chanae Group A isolation valves are in process lines that communicate directly with the reactor vessel.and penetrate the primary l containment. These lines have two isolation valves in neries,

! one inside and one outside the primary containment. The control system for Group A isolation valves is designed to:

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- provide closure of the valve in time to prevent uncovering the core as a result of a break in the line which the valves isolates; and

- provide closure of the valves with sufficient rapidity to restrict the release of radioactive material to the environs well below the values of 10 CFR 100.

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Isolation signals "B, C, D, E, & P" are part of an isolation function that generate the closure of various Group A components. Total system isolation (Group A) is not always desirable and for certain trip conditions only certain isolation functions are initiated.

Group A isolation functior.s include generation of isolation signals for the following components:

1. Main steam line isolation valves (Penetration X-7A,B C.D)

2. Main steam line drain isolation valves (Penetration X-8)

3. Reactor water sample isolation valves (Penetration X-41)

4. Condenser vacuum pump The condenser vacuum pump is tripped only for a main steam high radiation trip. Reactor water sampic isolation valves

'(02-A0V-39 & 40) are designed to receive an isolation signal to close for the following trip setpoints:

SIGNAL DESCRIPTION "B" Reactor vessel low low water level trip; and "C" Main steam line high radiation trip.

The remaining Group A isolation valves are designed to receive an isolation signal to close when trip setpoints are exceeded for any of the following parameters:

SIGMAL DESCRIPTION "B" Reactor vessel low low water level trip; "C" Main steam line high radiation trip; "D" Steam line high flow; "E" Steam tunnel high temperature; and "P" Main steam line low pressure.

The following paragraphs are a description of isolation signals "D, E, & P", which are not intended to cause reactor water l

j sample line isolation:

Nhin Steam Line Mich Flow ("D" Isolation Sinnal)

A main steam line high flow could indicate a break in a main steam line. The automatic closure of various i Group A valves prevents the excessive loss of reactor coolant and the release of significant amounts of radioactive material from the Reactor coolant Pressure Boundary. Upon detection of the main steam line high flow the following lines are isolated:

j f a. All four main steam lines (Penetration X-7A,5,C.D)

b. Main steam line drain (Penetration X-8)

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7 Main Steam Line SDSCe Mich TemDerature ("E" Isolation Signal)

A high temperature in the space in which the main steam lines are located outside of the primary containment could indicate a breach in a main steam line. The automatic closure of various Group A valves i

prevents the excessive loss of reactor coolant and the release of significant amounts of radioactive material 1

from the Reactor Coolant Pressure Boundary. When high temperatures occur in the main steam line space, the following pipelines are isolated:

a. All four main steam lines (Penetration X-7A.B.C.D)

b. Main steam line drain (Penetration X-8)

Low Steam Pressure at Turbinc Inlet ("P" Isolation

! Signal)

I Low steam pressure upstream of the turbine stop valves while the reactor is operating could indicate a malfunction of the pressure regulator in which the turbine control valves or turbine bypass valves open fully. This action could cause rapid depressurization of the reactor coolant system. From part-load operating conditions, the rate of decrease of reactor coolant system saturation temperature could exceed the design rate of change of vessel temperature. A rapid depressurization of the reactor vessel while the reactor is near full power could result in undesirable differential pressures across the channels around some fuel bundles of sufficient magnitude to cause mechanical deformation of channel walls. Such i

depressurizations, without adequate preventive action, could require thorough vessel analysis or core inspection before returning the reactor to power operation. To avoid the time-consuming requirements following a rapid depressurization, the steam pressure at the turbine inlet is monitored. Upon falling below a preselected value with the reactor in the RUN mode a

! "P" isolation signal initiates the isolation of the following lines:

a. All four main steam lines (Penetration X-7A,B,C,D)

b. Main steam drain line (Penetration X-0) l i Clearly, isolation signals "D, E & P" are from sensors that -

detect possible leakage outside the primary containment. .

Therefore, isolating the 3/4" size reactor water sample line (Penetration X-41) is pointless since no significant amount of reactor coolant inventory could escape; and release of radioactive material to the environs are well below the values 4

setforth by 10 CFR 100.

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In addition, isolation of main steam lines and main steam drain line due to signals "D, E. & P" results in a need for reactor coolant sampling analysis before returning the reactor to power operation.

The proposed changes to the Technical Specifications do not change any system or subsystem and will not alter the conclusions of either the FSAR or SER accident analyses.

i .The commission has provided guidance concerning the application l of the standards for making a "no significant hazard considerations" determination by providing certain examples in the Federal Register (F.R.) Vol. 48, No. 67 dated April i 6, 1984, page 14870. The proposed changes to Table 3.7-1 of

] the Technical Specifications match Commission example (1): "A i purely administrative change to technical specifications: for i

example, ... correction of an error, .... "

l Operation of the FitzPatrick plant in accordance with the

+, proposed amendments, therefore, would not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated since isolation signals "D.E, & P", as described above, are only appropriate for main steam line breaks, and closure of the reactor water sample valves will have no effect on this accident: or

! (2) create the possibility of a new or different kind of accident from any accident previously evaluated since isolation signals "D,E, & P", as described above, are not intended to isolate the reactor water sample line.

j Original design only called for signals "B & C"; or (3) involve a significant reduction in a margin of safety l since the signals mentioned above are inappropriate  ;

for reactor water sample line isolation and the same level of safety will still exist.

IV. Innlementation of the chances ,

Implementation of the changes, as proposed, will not impact the ALARA or fire protection program at Fit: Patrick, nor will the changes impact the environment. ,

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6 V. Conclusion The incorporation of these changes:

a) will not change the probability nor the consequences of an accidene or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report; I

b) will not increase the possibility or an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report; i

c) will not reduce the margin of safety as defined in the basis for any Technical Specifications; d) does not constitute an unreviewed safety as defined in 10 CFR 50.59; and e) involves no Significant Hazard Considerations, as defined in 10 CFR 50.92.

VI. References

a. James A. FitzPatrick Nuclear Power Plant Final Safety Analysis Report (FSAR), Rev. 2 July, 1984, Sections 7.2, 7.3, and 14.5.

b. James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER).

c. PASNY letter, J. P. Bayne to D. B. Vassallo, dated May 13, 1983 (JPN-83-42).

d. NRC letter, D. B. Vassallo to J. P. Bayne, dated February 21, 1984.

e. NYPA letter, J. P. Bayne to D. B. Vassallo, dated March 20, 1984 (JPN-84-18).

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