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Application for Amend to License NPF-47,revising Requirements for Tech Spec 3.5.2, ECCS-Shutdown.
	ML20043B230
Person / Time
Site:	River Bend
Issue date:	05/21/1990
From:	Plunkett T; GULF STATES UTILITIES CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20043B231	List: ML20043B230; ML20043B234;
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	RBG-32865, NUDOCS 9005250117
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x GULF STAX'X S UTXLITIES COMPANY i

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May 21,1990 i PBG- 32865 Fi.le No. G9.5, G9.42- I U. S.'. Nuclear Regulatory Cconission Docunent Control Desk ,j l Washington, D. C. 205S5 ~ Gentiment . j River Bend Station - Unit 1 Docket No. 50-458 j Gulf States 1 Utilities (GSU) Cmpany hereby files an application j . Station

                                 -to amend the River                       Bend                          -

Unit 1 Technical-Specifications,. Appendix A to Facility Operating License NPF-47,. i pursuant to 10CFR50.90. This application is filed to revise thes ; irequirements :for Specification 3. 5.2, . "ECCS-Ghutdown". This: 7 revision will allcw one loop of RHR to be used as an'ECCS- systs and also be used for shutdown cooling. This change is shnilar to

                                  -others approved for IEP.s including Grand Gulf ' (BNR/6) . . This proposed change'isirequested prior to the_ third refueling outage                                          j at River ' Bend Station. The~ attachment to this letter and                                              .l enclosure provide the -}ustif.ications' and' proposal revisions to the technical specifications.                                                                                     ;

Your prmpt attention to this matte:: is appreciated. .

                                                                                                                                              ~l t

Sincerely, N T. F. Plunkett General; Manager, Buslhess ;

     &                                                                                        Systems and Oversight                                   '

River Bend Nuclear Group 4 Attachment i- (( 9005250117 000521 m PDR 'ADOCK 05000458 P PDC , Q[ C. sn n 9 o;

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cc: U. S. Nuclear Regulatory Ccrimission b ', 611-Ryan Plaza Drive, Suite 1000 Arlingtory TX 76011

                                         . IFC Resident Inspector.-

Post Office Box 1051 St. Francisville, IA 70775 Mr. Walt-Paulson

              ,                            U.LS.' Nuclear Regulatory Ccrimission g
                                          .One White Plint North
                                          -11555 Rockville Pike Rockville, MD 20852 11r. William 11. Spell, Administrator Nuclear-Paergy' Division louisiana Dept. Of Dwironmental Quality P. O.: Box 14690                                     -{

Baton ' :. , IA 70898 1

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                                   . STATE OF. LOUISIANA                                      )

h's PARISH OF WEST FELICIANA~ ) .t r Docket No. 50-458

      ,                             In the Matter of                                          )

f ?> > GULF STATES. UTILITIES COMPANY )

                                       .(River Bend Station:- Unit 1)                                                                     I. A v

AFFIDAVIT. I T. F. Plunkett,' being duly sworn, states that he is a General ~ Manager - Business. Systems and Oversight Lfo Gulf States Utilities Company; that.he-is authorized on the part-;of' said" company to sign and file with the Nuclear Regulat'ory. Commission. the documents attached hereto; and that all such-

documents-are-true and correct to the best of his ' knowledge,. ,

_t l Linformation and? belief. 'f Q/f T. 'F. Pluhkett-k a A Subscribed.-and sworn to before me, a Notary Public in and (for the-State and Parish above named, this d/N day -of M Gli , 1990 . My Commission expires'with Life. j G y '. OhinAit 1 M Claudia F. Hurst , a i Notary -Public in aridifor;; ' y; West Feliciana Parish," Louisiana-. j3 . .. s i Du; h,. , '

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                   - Licensing Document Imolved:                      'Ibchnical Specifications -

Tint 9: i 3/4.3.2 PAGE: 3/4 3-13,15,18,26,27,29' 3.5.2 3/4 5-6 }

 \,                                                                                                    i RFASON POR BHQUEST
                    'In 'accordance- with 10CFR50.90 a revision to the River ~ Bend Station (RBS)

Unit l' Technical ~ Specifications, Appendix A to Facility Operating . License -[ NPF-47 .is: being. requested. Technical Specification 3.5.2, ECCS-Shutda n, ' requires 7that at least two DCCS loops be operable and capable of t transferring . water from the suppretsion pool to the reactor vessel. mis request asks that a note be added to technical specifications which~ allows one of. the ECCS ' loops to be a UCI subsystem aligned in she shutdown - cooling mode. C m pliance with the'above. technical specification has result = h' scheduled *

                  - outage maintenance and testing being delayed when unforesea m d.lems have         "

prevented an ECCS loop ' frcxn being returned to service. A change to technical specifications is requested to allow more flexibility to perform

                   - scheduled outage work and to reduce the length of subsequent outages while      i still maintaining an acceptable level of safety as described below.-            4 DESCRIPTIN This change request is the result of a review of the recent- successful outage ~ at' Grand Gulf. One factor in the success of the Grand Gulf outage
             ,       was:a provision in their technical ' specifications, ECCS-Shutdown, which 3_               allows :for the manual realignment of a LPCI subsystem in the shutdown cooling node to meet the operability requirement for an injection system.
                   , An . industry- review has found that the same provision exists in the             +

technical' specifications of IaSalle and Susquehanna. This change request asks that':the same provision be added to the River- Bend Technical

                   ~ Specifications.

The purpose of ECCS is to provide protection against postulated loss.of coolant accidents (IDCA) caused by ruptures in the primary system piping. , A number of ECCS design requirements are specified in 10CFR50.46 the most- ; I -impop ant of which is the post-LOCA peak cladding temperature (PCT) .of. [ 2200 F. When the reactor is shutdown, ECCS provides a means of restoring .; Page 1 of 6

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reactor vessel inventory 1 in case the reactor vessel is accidentally drained. . The safety limit, given in Section 2.1.4 -of. the- technical-specifications ' requires that the reactor water level be maintained above the top of active irradiated fuel. Water coverage over ensures ;

                       .that                     the cladding. _ temperatures are well below                                                            the2200 the          fue[F limit.            j Specification 3.5.2 (ECCS-Shutdown) requires that at least two . ECCS loops                                                                                       '
                       -be operable.                               One loop is sufficient to provide for core flooding and i                       cooling following a postulated draindown event. .                                                                          The second loop is redundant and is required to satisfy single failure criteria,                                                                                                     j i

This : proposed change will allow only one of the RHR loops to be credited 1 for EOCS-Shutdown operability when -in the shutdown cooling mode. The ! second ECCS_ loop will rmain . fully operable and capable of autmatic 3

                         -initiation.- For the purpose of analysis,- the fully operable loop is                                                                                          ,j assumed to fall and the operator is required to realign the loop in                                                                                              l shutdown cooling to the LPCI mode before the reactor coolant level can                                                                                             4 decrease belcw the top of active fuel.                                                                                                                              I The. spectrum of breaks and mishaps which have the potential for draining the reactor vessel can be divided into two categories, I) those which occur outside' containment, and II) those which occur inside containment.

I- Outside Containment Industry experience has found that the most probable cause of accidental drainage of the reactor vessel is the misalig ment -of the RHR valves. In 1985, such an event occurred at River Bend. The mishap was attributed to procedural deficiencies ' cmbined with operator -error (Reference LER 85-008). To prevent reoccurrence, ,

                                     ' interlocks were installed on the RHR pump suction valves                                                                              and precaution statennnts were added to the RHR operating procedure.                                                                                     1 Outside containment, the potential.for draining the reactor vessel is.                                                                                 !

limited' to the MIR and RWCU systems. RHR and RWCU are the only systems : which have lines beginning below the top of active fuel (TAF) and that ; exit containment. Both RHR and RWCU have divisional inboard and i outboard containrtentiisolation valves which are designed to close nn a '! low reactor water level signal. For all postulated . accidents,- the i containment isolation valves will close long before reactor water level can decrease to the top of active fuel (less than 1 minute). ; II Inside Containment i Inside containment, the potential for draining the reactor vessel is limited to small breaks and maintenance mishaps. Large line breaks J are not postulated to occur when the reactor is in cold shutdown and depressurized. (Reference USAR Section 15A.6.5.3) In a draindown event, the control rom operator is assumed to respond-to a loss of reactor coolant by the time reactor water level decreases to the RPS scram /RHR isolation setpoint, Invel 3 (+9.7") . Water level ~ is normally maintained at or above Level 5 (+35") when the reactor is ! Page 2 of 6 I b I - ^_ _ _ _ _ . _ _ _ _ - . ___.__m________m ___.______m_m._.m__ _ _ _ _ _ _ . _ _ _ _

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                                                           > shutdown rJ actual' operator- response is expected at the low level Jalnm setpoint, Ievel 4 (+30.5") . The time required for an operator to realign a 1KI subsystem frm the shutdown cooling mode _ to the LPCI s

injection - 's less than 7_' minutes. Ub ensure operator action, steps; , added to the appropriate procedures which instruct the align the IKI subsysts at level 3 (+9.7") . Maintenance , - operator E

                                                           -activities which have a- potential for draining'the reactor vessel                 y

= include a) instrument line breaks or valving errors, b)changeout ; of - i incore instrument drytubes, c) maintenance of reactor recirculation c" loop cmponents, and d) control rod drive maintenance. ..! q

 -.      ,                                                  a. All reactor vessel instrument and sample-lines are one inch in-                 !

diameter or smaller. For' all postulated -breaks and valving mishaps,_ the time required to drain the reactor vessel frm Level 3 to the top of active fuel is approximately 195 minutes.. , 1

b. Incore instrument drytubes are removed from the core by operators

                                                                 -on the refueling bridge.         The reactor vessel cavity .must be              >

j filled with water to provide shielding. To prevent: draining the 1 1 reactor, a - " water can" is installed underneath the reactor d ~ vessel. . Assuming a worst case scenario in which the drytube is j

renoved without a." water can" in place, the time required to l

drain the . vessel frm Level 3 -to the top of active fuel is : approximately 49 minutes,

c. Each reactor recirculation loop consists of ' a pump, a flow control valve, a suction isolation valve, and a discharge :

isolation valve. For maintenance of the pump or flow' control valve,'the suction and discharge _ isolation. valves are-assumed to. be closed with the valve operators electrically' disabled. Suction j and _ discharge ' isolation valves sometimes require the valve' stem y packing to be replaced. For this evolution, the: isolation valve N

                                                                  .will be backseated and the valve operator will be electrically.

disabled. Should the backseat fail, the time required to drain the, vessel frm level _3 to the top -of active ' fuel is ;

                ,                                                   approximately-32 minutes.                                                          !

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d. As a limiting event an- error in performing control rod. 1 maintenance was assumed. Control rod drives ' (CRDs)' are removed J frm undernectl1 the reactor vessel. The associated control rod blade backsents onto the guide tube to prevent the reactor vessel from draining. With the assumptions of 1, the CRD changeout is: l [l performed with reactor water level at Level 5 (+35")' andx2, the i control rod is unseated, the drain time from Level _3 (RHR- d isolation) to the top of . active fuel is approximately 28 minutes. l K The cmbination of both of these assumptions is unlikely because ,

CRD changeout is normally performed with 'the reactor cavity l' filled and technical specifications prohibits the movement of fuel ass ablies or control' rods at the lower water level, j . The- offsite c' io cor ~r x s ,postulateo draindown of the reactor ,

                                        -vessel arc not inc2 # v                         e     ,        alt of this change to technical of 6 r
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    . :. ~                                                                                                       l specifications.      A_ combination of ECCS autcmatic actuation, containment isolation, and prmpt operator action all assure that the reactor water-level is maintained above the top of active fuel. Fuel cladding failure is-precluded as:long as the core rcmains covered.       In addition, the . offsite ~

dose consequences are further mitigated by the technical'epecification

              - mquirement for primary containment integrity to De main +a.tned during . fuel handling, -core alterations, and operations with a potential for draining the reactor vessel. Therefore the radioactive release frm subsystes and
              'cmponents -discussed in Section 15.7 of the USAR are not affected by this

change to technical specificadions. Also, the individual- and' cumulative Eoccupational exposuras are not increased as a result of this change because the reactor water during a normal refueling outage is not affected.

The present technical specifications do not requim the RHR or RWCU isolation instruments _ to be operable while the reactor is shutdown, River Bend proposes to maintain at least one isolation division operable. This 3 will_ be accmplished by including a requirment to provide at least one division of the necessary RHR or RWCU level isole. ions or isolate- the affected system when depending on an RHR loop to be reconfigured. By j including this information in the note to specifications 3.3.2 (Isolation ' ' Actuation - ~ Instrumentation) and 3.5.2 the necessary restrictions to the

                ~ initial. conditions of this evaluation will be included in the technical specifications thereby satisfying the intent of 10CFR50.36. Notes (c) and (d) of Table 4.3.2.1-1 are deleted since they have expired.                                   ]g     ,

Pr m the forgoing discussion, it can be concluded that the operator will have sufficient time to react and that no credible mishap will result in reactor water level decreasing below the top of active fuel. q SIGNIFICANP IRZARDS 00NSIIERATION j In accardance with the requirements of 10CFR50.92, the following discussion

is provided in support of the detennination that no significant hazards- are i created or increased by the changes proposed in this amendment request. j

1. No'significant increase in the probabilit/ or the consequences of _l an accident previously evaluated results frm the proposed change j because: ;

i The potential for draining the reactor vessel is - nore likely to ! occur whether one or two ECCS systems are aligne. for autcmatic l initiation. The consequences of a draindown event are not increased as l'ong as water is maintained above the top of active fuel. Reactor water !

                            . level above the top of active fuel ensures adequate fuel cooling.
                                                                                                                   ~

A ECCS operability during a draindown event was evaluated assurning a failure of the ECCS loop aligned for autmatic actuation. The ] g control roam operator was found to have sufficient time to realign the other ECCS loop frm the shutdown ecoling mode to the IPCI al ; Page 4 of 6 ! l t %%. 4 M' /

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 ..                                                                                                                 1 mode long before reactor water level could decrease to below the cop of active fuel. Large line breaks inside containment are not postulated to occur when the reactor is in cold shutdown and depressurized because system stresses will be well below that of the design.

2. 'the pmposed change will not create the possibility of a new or different kind of accident than previously evaluated because:

This proposed change is limited to the operability of ETS sysi. ems I when the reactor is shutdown. It does not involve a physical ! change to plant equipnent nor is it expected to introduce new failure nodes for important to safety equipnent. This change is 4 not applicable when the plant is operating, therefore, the ECCS { response to a design basis accident is not changed. 4 A manual realignment of the LPCI system from the shutdown cooling I made to the LPCI iniection node has been shown to be equivalent to an automatic actuation. As discussed above, the control rom l operator has sufficient time to respond to a loss of reactor coolant for all mishaps considered.

3. This request would not invcive a significant reduction in the l margin of safety because:

i The bases of Technical Specification B3/4.5.2 describes ECCS as a { source of flooding the core in case of accidental draining. The i reactor vessel water level safety limit (Technical Specification l 2.1.4) was established at the top of active irradiated fuel to provide a point which can be monitored and also provide adequate i margin for effective action. If water level should drop below the ; top of active irradiated fuel, the ability to remove dacay heat is ] reduced and elevated cladding temperatures could result. This change to technical specifications relies on operator action to manually realign the shutdown cooling mde in case of a reactor vessel draindown event. Prmpt operator act-ion within 20 minutes assures ' hat no credible mishap will result in the reactor water - level decreasing below the top of active . irradiated fuel. "'he necessary instruments and autmatic actions relied upon in this evaluation have been required to be maintained operable when in this configuration. Manual realignment of RHR shutdown cooling will still provide for ECCS flood.tng of the core. Therefore, the margin of safety is not reduced. Because a cmbination of ECCS automatic actuation, containment isolation, and prmpt operator action all assure that the reactor water level is maintained above the top of active fuel, no fuel cladding failure is postulated. The dose consequences of a postulated draindown of the reactor vessel are not, thus, increased as a result of this change to the technical specifications. Radioactive release frm subsystems and cmponents as Page 5 of 6 1 l

s discussed in Sectics 15.7 of the USAR are not increased as a result of this W change to' technical specifications. In conclusion, the proposed change does not increase the possibility-or the . consequences of a previously evaluated accident, does not create a new or ! differenti kind of accident frm any previously evalua+.ed, and does not l involve a significant reduction in the margin of safety. GSU proposes that j no significant hazards considerations are involved. '

                               ' RINISED 'mCINICAL SPECIFICATION                                                                                                                                                              !

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                                  'Ihc requested revision is provided in the Enclosure.-                                                                                                                                   1 J

SCHEDUIE FOR ATTAINING COMPIJANCE ! As indicated above, RBS is currently in canpliance with the applicabic technicalispecifications. However, to utilize this change in scheduling , the third refueling outage, GSU requests that- the proposed change be l approved by August 30, 1990. This will allow _ adequate advanced planning i prior to the refueling outage, currently scheduled to begin September 15,

                                 -1990.

HDf!FICATION OF STATE PERSONNEL q A copy of the amendment application and this sub:Littal has been provided to : the State of Iouisiana, Department' of Environmental Quality - Nuclear j Energy Division. _{; ENVIRONIENmL IMPACP APPRAISAL l Gulf States Utilities Company '(GSU) has reviewed the proposed technical spwification change. against the criteria of 10CFR51.22 for environmental considerations.- As shown above, the proposed change -does not involve a significant hazard consideration, nor increase the types and amounts of effluents that may be released offsite, nor significantly -increase 1

                                 -individual or cumulative occupational radiation exposures. Based on the                                                                                                                   y foregoing, GSU concludes that the proposed technical' specification change                                                                                                              -

1 meets the criteria.given in 10CFR51.22(c) (9) for categorical exclusion from j

                                  . the requirement for an Environmental Impact Statement.                                                                                                                                          j 1

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E t_- C ! INSERIS J I Add the following notes to specification 3/4.3.2 'l I. Table 3.3.2-1 items 4.e and 6.c ' APPLICABLE OPERATIONATi CONDITION' and-h ,

                                                                 ' MINIMUM OPERABLE CHANPELS PER TRIP SYSTEM' as shown-on the r e kup
                                                                                  -(m). Also when the associated division is required under         'j specification 3.5.2 note'#'                                    i
                                                                                                                                                    .1 II.                         Table 4.3.2.3-1 -itens 4.e and 6.c ' APPLICABLE OPERATIONAL CONDITICN'                    l c                                                                as shcun on the markup                                                                    !

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                                                                                  .(c) Also when the associated division is required under specification 3.5.2 note'#'                             ^l
                                                                                                                                                    'I
                                                                                             .                                                         - l.

Add the-following note to speci" ation 3.5.2 l

                                   -III # One LPCI subsystem may be aligned in the shutdown cooling node .                                                 I
                                                              ..provided at least one division of the necessary RHR and RWCU level
                                                               -isolations are operable per specification 3/4.3.2.                                          ,

1 i i i = 1 - ii ~ i M

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ML20043B230

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