ML20012F220

Proposed Tech Specs Revising Section 3.5.F & 4.5.F Re Min ECCS Cold Shutdown Requirements

ML20012F220
Person / Time
Site:	FitzPatrick
Issue date:	04/02/1990
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
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ML20012F217	List: ML20012F216 ML20012F218 ML20012F220
References
NUDOCS 9004100415
Download: ML20012F220 (19)

Text

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ATTACHMENT l 1

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f PROPOSED TECHNICAL SPECINCATION CHANGE j REGARDING MINIMUM ECCS COLD SHUTDOWN ~

REQUIREMENTS (JPTS 88-002)

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l New York Power Authority

JAMES A. FITZPATRICK NUCLEAR POWER Pl. ANT Docket No. 50333 DPR 59

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N TABLE OF CONTENTS (Cont'd)

F. ECCS-Cold Condition F.

G. Maintenance of Filled Discharge Pipe G. 122al H. Average Planar Unear Heat Generation Rate (APLHGR) H. 123 1 1. Unear Heat Generation Rate (LHGR) 1. 124 J. Tnermal Hyd sufic Stability J. 124a

' SURVEILLANCE

! UMITING CONDITIONS FOR OPERATION REQUIREMENTS 3.6 Reactor Coolant System 4.6 136 A. Pressurization and Thermal Limits A. 136 B. DELETED i C. Coolant Chemistry C. 139 D. Coolant Leakage D. 141

. E. Safety and Safety / Relief Valves E. 142a F. Structuralintegrity F. 144 G. Jet Pumps G. 144 H. DELETED 1

1. Shock Suppressors (Snubbers) 1. 145b 3.7 Containment Systems 4.7 165 A. Primary Containment A. 165 B. Standby Gas Treatment System B. 181 i C. Secondary Containment C. 184 D. Primary Containment isolation Valves D. 185 3.8 Miscellaneous Radioactive Material Sources 4.8 214 3.9 Auxiliary Electrical Systems 4.9 215 A. Normal and Reserve AC Power Systems A. 215 B. Emergency AC Power System B. 216 C. Diesel Fuel C. 218 D. Diesel Generator Operability D. 220 E. Station Batteries E. 221 F. LPCI MOVIndependent Power Supplies F. 222a G. Reactor Protection System Electrical Protection Assemblies G. 222c 3.10 Core Alterations 4.10 227 A. Refueling interlocks A. 227 B. Core Monitoring B. 230 C. Spent Fuel Storage Poo1 Water Level C. 231 D. Control Rod and Control Rod Drive Maintenance D. 231 3.11 Additional Safety Related Plant Capabilities 4.11 237 A. Main Control Room Ventilation A. 237 B. Crescent Atsa Ventilation B. 239 i

C. Battery Room Ventilation C. 239 1

1 Amendment No, f,f, p,1[

JAFNPP .

3.5 (cont'd) 45 (cont'd)

F. ECCS Cold C6.et;cn F. ECCSCold Condibon e pressure systerns requwed

1. A mamum W two low pressure Emergency Core Cooling by 35.F.1 and 35.F2 sher be as folkms-Wems M be wm h'nm W's'm the reactor, the reactor is in the cold condition, and work is

1. Perf rm a flowrate test at least once overy 3 mones on me wim M W W dr m N r e requeed Core Spray pump (s) and/or the RHR pun 9(s).

Each Core Spray pump sher desver at least 4.625 gpm

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2. A momum d one low pressure En gancy Core Cooling i

subsystem shall be operable i c.; irradiated fuel is in 7e7 en a 2113 W h pnrnery E pump at the reactor, the reactor is in the cold condibon, and no gp agens a @ W 6 2 a work is bemg po.h. d with the potenbal for dranng the eg reactor vessel to pnmery contanment dlierentW pressure d > 20 psid.

3 Emergency Core Cooling subsysb.ns are not requwed to 2. Perform a monthly operability test on me requirsd Core '

be operable prowded that the reactor vessel head is Spray and/or LPCI motor operated valves.

r me % s , me W W W gates

. 3. Once each shlet verify the suppression pool water level is accordance withD

' " 88 8*

greater then or equel to 10.33 ft. whenever the low W, Y.3.10.C.

presstre ECCS subsystems are aliped to me appreemon

4. With the requwements d 35 F.1,35.F2, or 35.F.3 not sakssied, suspend core afterabons and as operabons with 4. Once each shift verify a mamum W 32Cmches d water is i the potenbal for dranng the reactor vessel. Restore at available in the Condensate Storage Tanks (CST) least one system to operable status vnthe 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or whenever the Core Spray System (s) is angned to me establish Sewnda y Contamment integnty vnthm the next tanks.

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Amendment No. [ t 122 i

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3.5 BASES (cont'd)

$ vessel head off the LPCI and Core Spray Systems will i mb.. F. ECCS-Shutdown Mode s their Wied safety funchon without the help d the ADS.

Low pressure W Core W Sysdoms A are E. Reactor Core Isolabon Cooling (RCIC) System requwed when me reactor is in a cold cont 9 Eon to ensuie C*** # ""

The RCIC is dessgned to provide maket, to the Reactor Coolant System as a planned operatior; for penods when the drandown et the reactor veessi. Two low pressure ECCS normal heat sink is unavailable. The RCIC also serves as are W we b M me @ ,

redundant makeup system on total loss of all offsde power in the event that HPCI is unavailable. In aR other postulated The low pressure ECCS WJ_. coneet of two CS acodents and transsents, the ADS prowsdes redundancy for the systems, two LPCI subsystems, or a comtanasion moreof.

HPCI. Based on this and judgements on the reliability of the Each CS system conoots of one motor-drwen pung, HPCI system, an allowable repar time of 7 days is specified. associated pipeg, and valves. Each CS system is capahle at Immediate and daily venficahons of HPCI operability dunng transiemng water to me rear *w veneel from the suppreseson RCIC outage is conssdered adequate based on judgement and pool or, when me sippreseson pool is unsweilable, the prachcarey. condensate storage tank. In the cold contShen, em:h LPCI subsystem consets of one motor <Mwen pump. W m power physses W and reactor operator training wth inoperable components wiR be conducted only when the RCIC N# #

System is not requwed, (reactor coolant temperature <212'F and coolant pressure <150 psig). If the plant parameiews are one RHR pump b W p W 6 i below the poet where the RCIC System is requred,Yp As because d Ms langer sowrate mmpmed e a Core Spray A operatng in me N W teshng and operator L ;.a* will not place the plant in an g mode of RHR is considered operable for me ECCS functon it it can be reaEgned manuesy (either remose or locaq to the LPCI Operability of the RCIC System is requwed only when reactor mode and is not othennse inoperable. In me cold cont 9 ton, pressure is gresler than 150 psig and reactor coolant the RHR system cross-tie valves are not requwed to be closed.

temperature is greater than 212"F because core spray and low pressure coolant infechon can protect the core for any size pipe break atlow pressure. tiooding capabildy to recover from an snedvertent vessel

• dr;endown. However, with only one low pressure system operable, the overall system reliabildy is reduced because a sogle-tailure could render the ECCS sncapable of p bT-g itsintended An.eninTeent No. f,If,If 129

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3.5 BASES (cont'd) funcbon. Therefore, ofm.aison with the potenhal for drarung generation rate of all the rods of a fuel assembly at any seisi the reactor vessel is not allcwed with only one low pressure locahon and is only dependent secondarily on wie rod to rod ECCS subsystem operable. power dstribubon wdhwi an samambly. Since emptW locai vanahons in power detribubon wehen a fuel assembly a:.act Wie condihons. Sullicsont coolant s ~=vry is available above the calculated peak dad temperature by less Wien + 20'F reistve fuel to apow operator achon to termnate the inventory loss b Wie peak 6 for a W fuel 4 Wie limit on .

prior to fuel ti.cc;;; i ni case of an inadvertent draridown. " *9' " 8"

the calculated hamperatures are witun Wie 10 CFR 50 G. Mantenance of Fi5ed Drscharge Pipe Appendix K limit. The limitng values for APLHGR are gwen in N the dscharge ppng of the core spray, LPCI, RCIC, and HPCI Figem 3M1 #wagh 3M4 W Ernihng values of are not filled, a water toi..s can h,"cy in this pipsng when as a W N W are gwen h NN-2 the pump (s) are started. To nunmze (-cays to the discharge (as W W W 6 M w.mWhd pipsng and to ensure added margin in the operation of these as a h d W and W W are p h systems, this technical w_ ,mo_ _. on requires m_-__ the u.m NEDC31317P (as amended) for Resood 7 and 8 fuel These lines to be fi5ed whenever the system is reqtiired to be " D MN operable. N a discharge pipe is not filled, the pumps the supply dM can be W h h M the line must be assumed to be inoperable for techrscal -

specir.caison purposes. However, if a water hammer were to I. Lsnear Heat Generation Rate (LHGR) m,the @1M M @ h % h This specificahon assures tiet Wie lineer hem generation rme in H. Average Planar Llaear Heat Generahon Rate (APLHGR) any rod is less than tie dessgn linear heet generatort This specdicahon assures that the peak clidi g temperature The LHGR shs5 be checked deley during rear *w operasson at follovnng the postulated desegn bases loss-of-coolant accsdent 25% rated thermai power to detemune N fuel bumup, or control will not exceed the limit speedied in 10 CFR 50 Appendix K. red n-(

. .;, has e mmad dienges in power detributort For The peak cicer s temperature sosowing a postulated low- LHGR b be a Emlhng value below 25% rated Wennel power, oodant w is pnmarily a function d ein w hem the ratio d local LHGR to average LHGR would have to be greater than 10 which is precluded by a considerable margin when employing any permssibie conard rod panerrt i

Amendn sit No. ,f,f 1 ,1[7 130

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JAFNPP ,

45 BASES (contd) the line is in a full condition. Between the monthly intervals at which the lines are vented, instrumentahon has oesn provided in the Core Spray System and IPCI System to monnor the presence of water in the 66d g. pipng. This instrumentabon will be calibrated on the same frequency as the salsty system instrumentabon. This period of penodic teshng emures that during the interval between the moneWy decks the status of the discharge pping is mondored on a conbnuous bases.

Normany the low pressure ECCS subsystems required by Specdicabon 35.F.1 are demonstrated operable by the surveillance tests in Specnicabons 45.A.1 and 45.A3. Sedon 45.F speedies penodic sunellance tests lor the low pressure ECCS subsystems which are applicable when the reactor is in the cold condibon. These tests in conjunchon with the requrements on Elled discharge piping (SpecNicaten 35.G),

and the requrements on ECCS actuabon instrumentabon (Speedicabon 3.2.B), assure ariary* ECCS capabitty in the cold condibon. The water level in the suppression pool, or the Condensase Storage Tanks (CST) when the siepression pool is -

inoperabie, is checked once each shut to ensure that suscient wateris available for core cooling.

133

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3.7 UMITING CONDITIONS FOR OPERATION 4.7 SURVEIUANCE REOUIREMENTS 1 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS WE4- Applicabiler.

Applies to the operating status of the pnmary and secondary Applies 10 the pnmary and secondary contamment inksgrity. ,

contamment ay-im Obrective- Obrectrve To assure the integnty of the p;cearf and secondary contanment To venty the wiegrey d the pnmary, and secondary contanment sydems. systems.

Sped w Specificanort A. Pnmary Contanment A. Pnmary Contanrrent

1. The volume and temperature of me water in the pressure 1. The pressure suppresson dumber water level and suppresson chamber shaN be mantaned vnthm the temperature sher be chodied once per day. The followng limes whenever the reactor is cnbcal or whewwer accessible intenor surfaces of me dryweR and above me the reactor coolant temperature is greeder than 212 F and water line of me pressure stepression chamber shen be irradiated fuel is in me reactor vessel- inspected at each retualing outage for owedence of detenorahon. Whenever there is "wwerman of relief valve

a. Maamum vent submergence lewei of 53 'mches. operabon or testag which adds host to be suppreseen

b. Mirumum vent submergence level of 51.5 inches.

e sheRbe W N and also observed and logged every 5 mmutes until the The stypression chamber water level may be heat addlbon is termmsted. Whenever more is indicahon outade the above limds for a maamum of four (4) of relief valve operation with me temperature of the hours dunng requwed operability testeg of HPCI, suppresson pool reachmg 19&F or more and me pnmay RCIC, RHR, CS, and the Stepresson Chamber - coolant system pressure greater then 200 poig, an ertemel DryweR Vacuum System. veuel exammehon of me appression chamber sher be

c. Masamumwater temperature c nduoed besore resurrung power operanon.

(1) During normal power operation maamum water temperature shall bc 95T.

Ar,einrgd No. [

165

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3 3.7 BASES (cont'd)

Using the minimum or maximum downcomer submergence Using a 4(TF rise (Saction 52 FSAR) in the suppression levels given in the specihcahon, contamment pressure dunng ct.asii w water temperature and a maximum irwhat temperature the design bases accident is approx;iTiately 45 psig which is d 957, a temperature of 145T is isGL;4 which is well below below the desigen of 56 psig. The midmtm downcomer the 1707 temperature which is used for complete -

subiTiergence of 51.5 in. results in a minimum suppression w iiu oin a v vii.

chamber water volorne of 105,600 ft3 . Tne majonty of the . .

Bodega tests (9) were run with a submerged length of 4 ft. and an rud - Wh e with ceivipiate condorsation. Thus, with respect to downcomer #

submergence, this specification is adequate. Addihonal pumps pumps W two M JAFNPP specific analyses done in curn.acticm with the Mark I pumps) N pre is M W W Containment-Suppression Ctam.s integrity Program indicate mamtam =- *wu=da net posstwo suchon head (HPSH) for the the adequacy of the specified range of submergence to ensure *** #8I P"*PS-4 that dynamic forces associated with pool swell do not result in 1.imshng suppression pool temperature to 1307 during RC'C, overstress d the suppression chamber or associated HPCs, or relier vasve operation, when decay heat and stored structures. energy are removed form the pnmary system by dischargmg The maximum tm.perature at the end of blowdown tested during the Humboldt Bay (10) and Bodega Bay tests was '"Y 1707, and ttN is conservahvely taken to be the limit for complete corduiir:,ation of the limit for complete condensabon Expenmental data indicates that excesswe steam condensmg of the teactor coolant, although condoi6aticm would occur for loads can be avoeded if the peck temperature of the temperatures above 1707. suppresseor; pool is maintamed below 1907 dunng any penod

! of relief valve operabon with sonic conditions at the disctwy .

exit. Specificativi6 have been placed on the envelope d reactor operating condibons so that the reactor can be depressurized in a tunely manne- to avoid the regime d potenhally high suppressson cha%er ica lings.

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r ATTACHMENT 11

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I SAFETY EVALUATION FOR THE PROPOSED i TECHNT5KEEPFciFICAff5FCRKR5TRE5KR51NG l

MINIMUM ECCS COLD SHUTDOWN REQUIREMENTS ,

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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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, Attachment 11 I SAFETY EVALUATION 4 Page 1 of 9

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l. DESCRIPTION OF THE PROPOSED CHANGE The proposed changes to the James A. FitzPatrick Technical Specifications revise Sections 3.5.F and 4.5.F, ' Minimum Emergency Core and Containment Cooling System Availability," on page 122. Two changes are proposed. The first change detes Specification 3.5.F.1 becauce it is redundant to Specificatiory M.A ar.J 3.5.B. The second change adds new Umiting Conditions  !

for Operation (LCC$) knv shopded Surveillance Requirements regarding ECCS availability with the reactor in tha @.o condutbn.

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! A. Eliminate Redundan. Umitino Condition to; Oteetlor)

Dalete existing Specification 3.5 F.1, on page 122:

"Any combination of inoperable components in the Core and Containment ,

l Cooling Systems shall not defeat the capability of the remaining operable  ;

components to fulfill the core and containment cooling functions.* .

B. New Specifications for ECCS Availability in Cold Condition  :

1. Replace existing Specification 3.5.F.2, on page 122, with the following:

3.5.F. ECCS-Shutdown Mode ,

1. A minimum of two low pressure Emergercy Core Cooling subsystems shall be operable whenever irradiated fuel is in the reactor, the reactor is in the cold condition, and work is being performed with the ootential for draining the reactor vessel. L

2. A minimum of one low pressure Emergency Core Cooling subsystem shall be operable whenever irradiated fuel is in the reactor, the reactor is in the cold condition, and no work is being performed with the potential for draining the reactor vessel.

3. Emergency Core Cooling subsystems are not required to be operable provided that the reactor vessel head is removed, the cavity is flooded, the spent fuel pool gates are removed, and the water level above the fuel is in accordance with Specification 3.10.C.  ;

4. With the requirements of 3.5.F.1,3.5.F.2, or 3.5.F.3 not satisfied, suspend .

core alterations and all operations with the potential for draining the reactor l vessel. Restore at least one system to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or i establish Secondary Containment Integrity within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

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Page 2 of 9 [

2. Replace existing Specification 4.5.F, on page 122, with the following:

4.5.F ECCS-Shutdown Mode Surveillance of the low pressure ECCS systems required by 3.5.F.1 and l l 3.5.F.2 shall be as follows:

1. Perform a flowrate test at least once every 3 months on the required l Core Spray pump (s) and/or the RHR pump (s). Each Core Spray pump shall deliver at least 4,625 gp:n against a system head corresponding to a reactor vessel pressure greater than or equal to  ;

t 13 psi above primary containment pressure. Each RHR pump ,

l shall deliver at least 9900 gpm aq,ainst a system head corresponding to a reactor vessel to prirnary containment differential pressure of > 20 paid, i i 2. Perform a monthly operability test on the required Core Spray and/or LPCI motor operated valves.

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3. Once each shift verify the suppression pool water level is greater than or equal to 10.33 ft, whenever the low pressure ECCS subsystems are aligned to the suppression pool.

4. Once each shift verify a minimum of 324 inches of water is available in the Condensate Storage Tanks (CST) whenever the Core Spray System (s) is aligned to the tanks.

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3. Shift existing Specification 3.5.G and 4.5.G, Maintenance of Filled Discharge Pipe, to l a new page numbered 122a.

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4. Replace existing Bases 3.5.F,on pages 129 and 130, with the following:

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F. ECCS Shutdown Mode -

Low pressure Emergency Core Cooling Systems (ECCS) are required when the reactor is in a cold condition to ensure adequate coolant inventory makeup in case of an inadvertent draindown of the reactor vessel. Two low pressure ECCS subsystems are required operable to meet the single-failure criterion.

The low pressure ECCS subsystems consist of two CS systems, two LPCI subsystems, or a combination thereof. Each CS system consists of one motor driven pump, associated piping, and valves. Each CS system is capable of transferring water to the reactor vessel from the suppression pool or, when the suppression pool Is unavailable, the condensate storage tank, in the cold condition, each LPCI subsystem consists of one motor driven pump, associated piping, and valves. Each LPCI cubsystem is capable of transferring water from the suppression pool to the reactor vessel. Only one RHR pump is required per LPCI subsystem because of its larger flowrate t

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l Attachment ll l SMETHVAUMTION '

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i compared to a Core Spray System. A LPCI subsystem operating in the shutdown cooling mode of RHR la considered operable for the ECCS I function if it can be realigned manually (either remote or local) to the LPCI  :

! mode and is not otherwise inoperable, in the cold condition, the RHR I l system cross-tie valves are not required to be closed. I One low pressure ECCS subsystem provides sufficient vessel flooding capability to recover from an inadvertent vessel draindown. However, with only one low pressure system operable, the overall system reliability is reduced because a single-failure could render the ECCS Incapable of i performing its intended function. Therefore, operation with the potential for '

draining the reactor vessel is not allowed with only one low pressure ECCS subsystem operable.

J ECCS systems are not required to be operable during refueling conditions. i Sufficient coolant inventory is available above the fuel to allow operator j action to terminate the inventory loss prior to fuel uncovery in case of an inadvertent draindown. ,

5. Add the following paragraph to the end of Bases Section 4.5, on page 133:

Normally the low pressure ECCS subsystems required by Specification 3.5.F.1 are demonstrated operable by the surveillance tests in Specifications 4.5.A.1 and 4.5.A.3. Section 4.5.F specifies periodic surveillance tests for the i Iow pressure ECCS subsystems which are applicable when the reactor is in the cold condition. These tests in conjunction with the requirements on filled discharge piping (Specification 3.5.G), and the requirements on ECCS actuation instrumentation (Specification 3.2.B), assure adequate ECCS

, capability in the cold condition. The water level in the suppression pool, or ,

the Condensate Storage Tanks (CST) when the suppression pool ls inoperable, is checked once each shift to ensure that sufficient water is >

l available for core coofing, ,

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6. Revise Specification 3.7.A.1, " Suppression Chamber," on page 165:

Delete the cross reference to Specification 3.5.F.2 and add the phrase i i

"whenever the reactor is critical or whenever the reactor coolant temperature is greater than 212 F and irradiated fuel is in the reactor vessel."

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7. Delete the following paragraph from Bases Section 3.7,

• Primary Containment", on page 188: j Should it be necessary to drain the suppression chamber, this should only be done when there is no requirement for Emergency Core Cooling Systems operability as explain 6d in basis 3.5.F.

8. Revise Table of Contents, on page 11:  ;

The titles of Sections 3.5.F and 4.5.F are changed to "ECCS-Cold Condition."

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11. PURPOSE OF THE PROPOSED CHANGE The purpose of these Technical Specification changes is to delete a duplicate specification on ECCS operability, and to introduce new LCOs and surveillance '

requirements for ECCS availability when the reactor is in the cold condition.

A. Eliminate Redundant Umitina Condition for Operation This proposed change deletes Specification 3.5.F.1 which is redundant to Specification 3.5.A and 3.5.B. Amendment 83, which incorporated the definition of " operable,"

introduced the redundant requirements for ECCS operability. The term " operable" when used in conjunction with Specifications 3.5 A and 3.5.B ensures that inoperable components do not defeat the capability of the ECCS and Containment Cooling Systems to fulfill their functions; thus, duplicating the ECCS operability requirements of

, Specification 3.5.F.1.

I B. New Specification for ECCS Avaliability in Cold Condition r

The issue of technical specification requirements for ECCS systems during outages was raised by the NRC during the January,1988 maintenance outage inspection at the Fitzpatrick plant (See inspection Report No. 88-01, Reference 3). The inspector noted:

1. that the Fitzpatrick plant's technical specifications are silent regarding ECCS i

operability for work which has the potential for draining the vessel; 1

2. that based en the availability of one core spray system and other non-ECCS systems to inject into the vessel, no technical safety concerns exist; l 3. that, administratively, the iequirements for ECCS systems during this condition should be defined more clearly.

l l The Authority agrees with this observation and proposed that these requirements be defined in the Technical Specifications. This proposed amendment revises Specifications 3.5.F and 4.5.F to require two low pressure ECCS systems to be operable whenever irradiated fuel is in the reactor, the reactor is in the cold condition, and work Is being performed which has the potential to drain the reactor vessel.

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l Attachment ll SAFETY EVALUATION Page 5 of 9 l

J The proposed Umiting Condition for Operation and Surveillance Requirements reflect current plant practices and are similar to the requirements established in the Standard .

Technical Specifications (Reference 1).

Analysis of ECCS Techrleal Specification Requirements The proposed Bases Section establishes that the core spray (CS) system and the low pressure coolant injection mode of the RHR system (LPCI) are the primary sources of ,

emergency core cooling in the event of an inadvertent draindown of the reactor vessel during cold shutdown conditions.

The consequences of an inadvertent dralndown of the reactor vesse! are bounded by the loss of coolant accident (LOCA). The long-term cooling analysis (References 4 and 10) following a design basis LOCA demonstrates that only one low pressure ECCS subsystem is required, post LOCA, to maintain the peak cladding temperature below the allowable -

limit. This analysis evaluated the entire spectrum of LOCA pipe break sizes. The limiting break size is the double-ended guillotine break of the recirculation suction line (4.17 ft2 )

which is, by definition, a larger opening than any opening associated with an inadvertent dralndown of the reactor vessel.

The proposed technical specifications require two low pressure ECCS subsystems to be operable whenever Irradiated fuel is in the reactor, the reactor is in the cold condition, and the potential exists for draining the reactor vessel. Two systems are required operable to satisfy single-failure criterion. Only a single RHR pump is required per LPCI subsystem because of its larger flowrate compared with a core spray system's flowrate.

! One low pressure ECCS subsystem provides sufficient vessel flooding capability to recover from an inadvertent vessel draindown. However, the overall system reliability is '

reduced because a single-failure in the system concurrent with a vessel draindown could result in the ECCS not being able to perform its function. Therefore, operation with the potential for draining the reactor vessel and thus uncovering the irradiated fuel are not -

allowed.

ECCS subsystems are not required during refueling conditions or during other operations when the reactor vessel head is removed, the head cavity is flooded, the spent fuel pool gates are removed, and approximately 23 feet of water Is maintained over the top of the reactor pressure vessel flange. The large inventory of water allows timely operator action to terminate an inadvertent draindown event prior to fuel uncovery. Therefore, core alterations and operations with the potential for dralning the reactor vessel are permitted.

in the event that no low pressure ECCS subsystems are operable and the water level requirements of Specification 3.10.C are not met, core alterations and operations with the potentla! for draining the reactor vessel will be suspended immediately. Timely restoration of emergency core cooling is required or Secondary Containment integrity is established to control potential releases of radioactivity.

Normally the low pressure ECCS subsystems required by Specification 3.5.F.1 are demonstrated operable by the surveillance tests in Specifications 4.5.A.1 and 4.5.A.3. The

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proposed technical specifications consolidate the surveillance tests which are applicable in the cold condition into one subsection. ,

in the cold condition, the pressure suppression function of the torus is not required.

However, the suppression pool is required to be operable as part of the low pressure

( ECCS systems. The water level in the suppression pool is checked once each shift to ensure sufficient inventory is available for core cooling. A minimum water level is specified based on NPSH, instrument inaccuracles, and the recirculation volume (i.e., the volume of water required to flood the bottom of the drywell shell up to the height of the vent pipe);

plus a safety margin of 25,000 gallons for conservatism. In addition, the proposed technica5 specifications clarify that the suppression chamber volume and temperature  :

l requirements (Specification 3.7.A.1) are not necessary in the cold condition.

Repair work might require making the suppression chamber inoperable (e.g., draining for l surface inspections). Specification 4.5.F.5 will permit these repairs to be made and at the  !

same time ensure that the irradiated fuel in the reactor vessel has an adequate cooling water supply.  !

I lil. IMPACT OF THE PROPOSED CHANGE A. Eliminate Redundant Umlting Condition for Operation The proposed change to delete Specification 3.5.F.1 is purely administrative in nature, it eliminates a redundant requirement for core and containment cooling operability. The proposed change does not involve modification of any existing equipment, systems, or components; nor does it relax any administrative controls or limitations imposed on existing plant equipment.

B. New Specifications for ECCS Availability in Cold Condition The changes which add new ECCS LCOs and surveillence requirements constitute an additional limitation on plant operations that are not presently included in the technical specifications. The new requirements are administrative in nature because they are ,

consistent with current plant policy and practice.

Operation of the plant in accordance with the proposed amendment is not a safety concern. The '

conclusions of the plant's accident analyses as documented in the FSAR or the NRC Staff's SER are not altered by these changes to the Technical Specifications.

IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the James A. FitzPatrick Nuclear Power Plant in accordance with the proposed amendment would not involve a significant hazards consideration as defined 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 proposed change to Specification 3.5.F.1 deletes a duplicate requirement for ECCS and Containment Cooling operability. The proposed change is purely administrative in nature and does not involve modification of any existing equipment, r

. Attachment ll  ;

SAFETY EVALUATION 1 Page 7 of 9 l l

systems, or components; nor does it relax any administrative controls or limitations i imposed on existing plant equipment. The change does not impact previously evaluated accidents; nor does M affect safe plant operations, l

The proposed changes which Introduce new specifications for ECCS availability in the cold j condition constitute an additional limitation beyond what is presently in the technical specifications. The new specifications require two low pressure ECCS subsystems to be operable whenever irradiated fuel is in the reactor, the reactor is in the cold condition, and the potential exists for draining the reactor vessel; thus ensuring adequate coolant inventory makeup in case of an inadvertont draindown of the reactor vessel. The consequences of an inadvertent draindown of the reactor vessel are bounded by the loss of coolant accident (LOCA) analysis. The proposed changes do not alter the conclusions of the plant's accident analyses as documented in the FSAR or the NRC's SER.

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

The proposed changes are administrative in naturc. They more clearly define the requirements for ECCS operability and ECCS availability. The changes do not involve modification to any of the plant's systems, equipment, or components; nor do they introduce any new failure modes. The proposed changes are consistent with current plant operating practices and do not allow plant operation in an unanalyzed configuration.

3. involve a significant reduction in the margin of safety. The proposed change to Specification 3.5.F.1 deletes a duplicate requirement for ECCS and Containment Cooling operability. The existing limiting conditions for operability (Sections 3.5.A and 3.5.B) and associated surveillance requirements (Sections 4.5.A and 4.5.B) are unchanged by this proposed amendment. The availability and operability requirements imposed on these >

systems are,thus, unchanged.

The proposed changes which introduce new specifications for ECC-S availability in the cold l

condition provide a slight increase in the margin of safety. These new specifications, which reflect current plant practices, formally prohibit operations with the potential for draining the t reactor vessel when coolant inventory makeup is not available. The changes do not involve l any plant modifications, nor do they affect the FSAR information regarding the emergency core cooling systems, in the April 6,1983 Federal Register (48FR14870), NRC published examples of license amendments that are not likely to involve a significant hazards consideration. Example (i) and (ii) from this Federal Register are applicable to the proposed changes:

"A purely administrative change to technical specifications: for example, a change to achieve consistency throughout the technical specifications, correction of an error, -

or a change in nomenclature."

and i "A change that constitutes an additional limitation, restriction, or control not presently includod in the technical specifications: for example, a more stringent surveillance requirement."

t

. Attachment 11 SAFETY EVALUATION Page 8 of 9  :

The proposed changes can be classified as not likely to involve significant hazards considerations, since the changes are purely administrative in nature and constitute an additional limitation on plant operations. The proposed amendment does not involve hardware changes nor any other changes to the plant's safety related structures, systems, or components V. IMPLEMENTATION OF THE PROPOSED CHANGE Implementation of the proposed change will not impact the ALARA or Fire Protection Programs at the FitzPatrick plant, nor will the change impact the environment.

VI. CONCLUSION The changes, as proposed, do not constitute an unreviewed safety question as defined in 10 CFR 50.59, That is, they:

a. will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report;

b. will not increase the possibility for an accident or malfunction of a type different from any l

evaluated previously in the safety analysis report; I

l c. will not reduce the margin of safety as defined in the basis for any technical specification; and

d. involves no significant hazards consideration, as defined in 10 CFR 50.92.

Vll. REFERENCES

1. NUREG 0123, Standard Technical Specifications for General Electric Bolling Water Reactors (GE STS), BWR/4.

2. James A. FitzPatrick Nuclear Power Plant Operating Procedure, F OP 13, Rev. 46, Residual i Heat Removal System.

3. USNRC Letter, dated March 29,1988, transmitting results of Inspection Report 50-333/88-01,

4. NEDC 31317P, SAFER /GESTR LOCA, October 1986, Loss-of Coolant Accident Analysis.

5. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report, Section 6.4,6.5,14.5.5, and 14.6.1.3.

6. USAEC " Safety Evaluation of the James A. FitzPatrick Nuclear Power Plant" (SER), dated November 20,1972.

- Attachment 11

, SAFETY EVALUATION Page 9 of 9

7. USAEC

• Supplement 1 to the Safety Evaluation of the James A. FitzPatrick Nuclear Power Plant" (SER), dated February 1,1973.

8. USAEC

• Supplement 2 to the Safety Evaluation of the James A. FitzPatrick Nuclear Power Plant * (SER), dated October 4,1974.

9. Amendment 83 to the James A. Fitzpatrick Operating Ucense, dated August 28,1984.

10. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report, Section 14.6.1.3

• Loss of Coolant Accident,' Table 14.61,

11. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report, Section 6.4.3

• Core Spray System,' Figure 6.4 2.

12. USNRC Letter, dated December 1,1989, transmitting results of Inspection Report 50-333/89 10.

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