ML20244A816

Proposed Tech Specs,Allowing B Loop of RHR HXs on Each Unit to Be Fed from RHR C & D Svc Water Pumps on Unit 1 Via cross-tie Line

ML20244A816
Person / Time
Site:	Quad Cities
Issue date:	03/31/1989
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20244A691	List: ML20244A689 ML20244A816
References
NUDOCS 8904180206
Download: ML20244A816 (16)

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4 ATTACIMEET_1 PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS FOR' OUAD CITIES UNITS 1 AND 2 DPR-29 DPR-30 I

3.5/4.5-4 3.5/4.5-3 3.5/4.5-17 3.5/4.5-11 l

8904180206 B90331 PDR ADOCK 05000254 P PNU 5590k s

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QUAD-CITIES

-DPR-29

! 1. b. From the' effective date ofg b. Flow rate test - After_ pump L this amendment until July , Agn/ /, /Ho each RHR service maintenance

', B&2, the % loop of thg._ as* ._

water pump shall and every

containment-cooling mode of deliver at least 3 months.

the RHR system for each n 3500 gpm against reactor may share the Unit e i a pressure of 198 EAH-end-@,RHR service c*4nd 'o' psig water pumps using cross tie line 1/2-10124-10" D.4 1/2.foroti6"-c. A logic system Each

Consequently, the require- o functional test refueling ments of Specifications outage 3.5.B.2 and 3.5.B.3 will impose the corresponding surveillance testing of equipment associated with

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'both react' ors'if~the'sha~ red RHR service water pump or pumps, or the crc ss tie line, are made or found to I be inoperable. T

2. From and after the date that one 2. When it is determined that.one of the RHR service water pumps RHR service water pump is:inop-is made or found to be inoper- erable, the remaining components able for any reason, continued of that loop and the other con-reactor operation is permissible tainment cooling loop of the RHR only during the succeeding 30 system shall be demonstrated to.

~ days unless such pump is sooner be operable immediately and  !

made operable, provided that daily thereafter. i during such 30 days all other active components of the con-tainment cooling mode of the RHR i system are operable. j

3. From and after the date that one 3. When one loop of the containment' I loop of the containment cooling cooling mode of the RHR system

. mode of the RHR system is made becomes inoperable, the operable .:

or found to be inoperable for loop shall be demonstrated to be '

any reason, continued reactor operable immediately, and daily operation is permissible only thereafter.

during the succeeding 7 days un-less such subsystem is sooner made operable, provided that all active components of the other

-loop of the containment cooling mode of the RHR system, both core spray subsystems, and both diesel generators required for operation of such components if no external source of power were available, shall be operable.

3.5/4.5-4 Amendment No. 114

QUAD-ClIIES

+

DPR-29 I

'The Containment Cooling mode of the RHR System consists of two loops.

Each loop consists of 1 Heat Exchanger, 2 RHR Pumps, and the associated valves, piping, electrical equipment, and instrumentation. The NFL +- 'A' gnfil April 1,1990,] Qoop on each unit contains 2 RHR Service Water Pumps. Dutiny ttTe i l pe14ed f= hvember 24, 1981, h e.' and

• o-] unit may utilizeW "A" and "C" RHR Service. Water Pumps from U

[After- April ,,,99 pia RHR a cross-tie Service linPAfter My 1,1982, each Wloop will contain 2g Water Pumps. Either set of equipment is capable of

. performing the containment cooling function. Loss of one RHR service water pump does not seriously jeopardize the containment cooling capability, as any one of the remaining three pumps can satisfy the cooling requirements. Since there is some redundancy left, a 30-day l repair period is adequate. Loss of one loop of the containment cooling l

mode of the RHR system leaves one remaining system to perform the containment cooling function. The operable system is demonstrated to be operable each day when the above condition occurs. Based on the h

fact that when one loop of the containment cooling mode of the RHR .h system becomes inoperable, only one system remains, which is tested daily, a 7-day repair period was specified.

O C. High-Pressure Coolant Injection

The high pressure coolant injection subsystem is provided to adequately cool the core for all pipe breaks smaller than those for which the LPCI mode of the RHR system or core spray subsystems can protect the core.

The HPCI meets this requirement without the use of offsite eiectrical power. For the pipe breaks for which the HPCI is intended to function, the core.never uncovers and is continuously cooled, thus no cladding damage occurs (reference SAR Section 6.2.5.3). The repair times for the limiting conditions of operation were set considering the use of the HPCI as part of the isolation cooling system. >

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D. Automatic Pressure Relief '

The relief valves of the automatic pressure relief subsystem are a backup to the HPCI subsystem. They enable the core spray subsystem and LPCI mode of the RHR system to provide protection against the small pipe break in the event of HPCI failure by depressurizing the reactor '

vessel rapidly enough to actuate the core spray subsystem and LPCI mode of the RHR system. The core spray subsystem and the LPCI mode of the RHR system provide sufficient flow of coolant to limit fuel cladding temperatures to less than 2200 F, to assure that core geometry remains intact, to limit the core wide clad metal-water reaction to less than 1%, and to limit the calculated local metal water reaction to less than 17%. ~

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3.5/4.5-17 Amendment No. 114 l

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' QUAD-CITIES

. DPR-30

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. continu:d reactor. c:ntainment cooling mode of the U operation is permissible' "

RHR shall be demonstrated to be ,

on1y during the succeeding 7. > operable immediately and daily j

, days.unless it is. sooner made thereafter.

operable, provided that during such 7 days all active compo- ' '

nents of both core spray sub-systems, the containment cooling mode of the RHR (including two l RHR pumps), and the diesel gen-- I erators . required for operation 1 of such components if no exter -

nal source of power were avall-able shall be operable.

. 6. If the requirements of Specifi-i cation 3.5.A cannot be met,'an orderly shutdown of the reactor shall be initiated, and the re- .!

actor shall be in the cold shut-

.down condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. Containment Cooling Mode of the RHR 8. Containment Cooling Mode of the RHR System System Surveillance of the containment cooling mode of the RHR system shall

'be performed as follows:

1. a. Both loops of the 1. RHR service water subsystem containment cooling mode of testing:

the RHR system, as defined in the bases for Spe- Item Frequency cification 3.5.8, shall be operable whenever irradiated a. Pump and valve Once/3 fuel is in the reactor operability . months vessel and prior to reactor startup from a cold condition.

1. b. From the effective date'of b. Flow rate test - After pump this amendment until each RHR service maintenance April 1,1990, water pump shall and every the "B" loop of the deliver at least 3 months containment ceding mode of . 3500 gpm against j the RHR system for each a pressure of 198 4 reactor may share the Unit 1 psig "C" and "D" RHR service

c. Nh waterpumpsusink" line 1/2-10509-1 crosstie D. Alogicsfstem functiona test ,alueling l

Consequently, the require- outage ments of Specifications 3.5.8.2 and 3.5.B.3 will impose the corresponding surveillance testing of equipment associated with both reactors if the shared ,

l RHR service water pump or i pumps, or the cross tie j line, are made or found to i be inoperable.

2. From and after the date that one 2. When it is determined that one f of the RHR service water pumps RHR service water pump is inop-(-

i is made or found to be Inoper- erable, the remaining components 3 j

' able for any reason, continued of that loop and the other con-  ;

reactor operation is permissible tainment cooling loop of the RHR .

anly during the succeeding 3') system shall be demonstrated to  !

iays unless such pump is sooner be operable immediately and j 4:ade operable, provided that daily thereafter, furing such 30 days all other active components of the con-tainment cooling mode of the RHR f system are operable, j 15458 3.5/4.5-3 Amendment No. 68, 72, 96 i

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QUAD-CITIES DPR-30 3.5 LIMITING CONDITIONS FOR OPERATION BASES A. Core Spray and LPCI Mode of the RHR System i

This specification assures that adequate emergency cooling capability i is available, i Based on the loss-of-coolant analyses included in References 1 and 2 and in accordance with 10 CFR 50.46 and Appendix K, core cooling systems provide sufficient cooling to the core to dissipate the energy associated with the loss-of-coolant accident, to limit the calculated peak cladding temperature to less than 2200'F, to assure that core geometry remains intact to limit the corewide cladding metal-water l reaction to less than 1% and to limit the calculated local metal-water I reaction to less than 17%.

The allowable repair times are established so that the average risk rate for repair would be no greater than the basic risk rate. The  !

method and concept are described in Reference 3. Using the results developed in this reference, the repair period is fcund to be less than half the test interval. This assumes that the core spray subsystems and LPCI constitute a one-out-of-two system; however, the combined effect of the two systems to limit excessive cladding temperature must 1 also be considered. The test interval specified in Specification 4.5 was 3 months. Therefore, an allowable repair period which maintains the basic risk considering single failures should be less than 30 days, and this specification is within this period. For multiple failures, a shorter interval is specified; to improve the assurance that the remaining systems will function, a daily test is called for. Although it is recognized that the information given in Reference 3 provides a quantitative method to estimate allowable repair times, the lack of operating data to support the analytical approach prevents complete acceptance of this method at this time. Therefore, the times stated in the specific items were established with due ' regard to judgment.

Should one core spray subsystem become inoperable, the remaining core spray subsystem and the entire LpCI mode of the RHR system are available should the need for core cooling arise. To assure that the  !

remaining core spray and the LPCI mode of the RHR system are available, they are demonstrated to be operable immediately. This demonstration includes a manual initiation of the pumps and associated valves. Based on judgments of the reliability of the remaining systems, i.e., the core spray and LPCI, a 7-day repair period was obtained.

Should the loss of one RHR pump occur, a nearly full complement of core ,

and containment cooling equipment is available. Three RHR pumps in i conjunction with the core spray subsystem will perform the core cooling function. Because of the availability of the majority of the core cooling equipment, which will be demonstrated to be operable, a 30-day repair period is justified. If the LPCI mode of the RHR system is not available, at least two RHR pumps must be available to fulfill the containment cooling function. The 7-day repair period is set on this basis.

B. RHR Service Water The containment cooling mode of the RHR system is provided to remove heat energy from the containment in the event of a loss-of-coolant accident. For the flow specified, the containment long-term pressure is limited to less than 8 psig and is therefore more than ample to provide the required heat-removal capability (reference SAR Section 5.2.3.2).

The Containment Cooling mode of the RHR System consists of two loops.

Each loop consists of I Heat Exchanger, 2 RHR Pumps, and the associated valves, piping, electrical equipment, and instrumentation. The "A" loop on each unit contains 2 RHR Service Water Pumps. Until April 1, 1990, the "B" loop on each unit may utilize the "C" and "0" RHR Service Water Pumps from Unit I via a cross-tie line. Af ter April 1, 1990, each "B" loop will contain 2 RHR Service Water Pumps. Either set of equipment is capable of performing the containment cooling function, loss of one KHR service water pump does not seriously jeopardize the containment cooling capability, as any one of the remaining three pumps can satisfy the cooling requirements. Since there is some redundancy left, a 30-day repair period is adequate. Loss of one loop of the containment cooling mode of the RHR system leaves one remaining system to perform the containment cooling function. The operable system is demonstrated to be operable each day when the above condition occurs.

15458 3.5/4.5-11 Amendment No. 68, 72, 96

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. AITAC1992fT 2 EtJMMARY OF CHANGES The following changes have been identified for Quad Cities Station Units 1 and 2:

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1) Page 3.5/4.5-4 Limiting Condition for Operation (LCO) 3.5.B.1.b (DPR-29)

Page 3.5/4.5-3.LCO 3.5.B.1.b (DPR-30)

(a) Change from "From the effective date of this snendment until July 1, 1982" to "From the effective date of this amendment until April 1, 1990..."  ;

(b) Change from the "A" loop to the "B" loop. _ ,

(c) Change from Unit 2 "A" and "B" RHR Service Water pump to Unit 1 "C" and "D" RHR Service Water pump.

(d) Change from using cross-tie line 1/2-10124-16"-D to cross-tie line 1/2-10509-16"-D.

2. Page 3.5/4.5-17 LCO Bases EHILSEVhe_tlater (DPR-29)

Page 3.5/4.5-11 LCO Bases RHR Service Water (DPR-30)

(a) Change from " November 24, 1981 to July 1, 1982" to "Until April 1, 1990".

(b) Change from "the "A" loop on each unit may utilize the "A" and "B" RHR Service Water Pumps f rom Unit 2 via a cross-tie line. After July 1, 1982, each "A" loop..." to "the "B" loop on each unit may utilize the "C" and "D" RHR Service Water Pumps from Unit 1 via cross-tie.

After April 1, 1990, each "B" loop..."

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3. Page 3.5/4.5-11 (DPR-30)

(a) Correct typographical error "RHR5" to "RHR".

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l NITACHMENT 3 t

DESCRIPTION AND BASES FOR AMENMiENT REOUEST i

A revision to the Quad Cities Units 1 and 2 Technical Specification is being proposed due to a crack which was discovered in the Unit 2 "B" loop of the RHR Servico Water piping located underground. This attachment provides a discussion of and bases for the proposed change.

DACEGEDUUD On Friday, March 10, 1989, excessive ground water leakage was observed in the tunnel which accesses the High Pressure Coolant Injection (HPCI) rooms. The HPCI tunnel is not a high traffic area and the abnormal leakage was observed during a non-routine tour. An investigation was conducted and determined that the leakage was due to a crack in the Unit 2 Residual Heat Removal (RHR) Service Water System "B" loop piping which is located underground. Previous surveillance conducted to demonstrate the system operability did not identify the leak since the leak was not large enough to impact the required system flow rates.

A similar leak was identified on November 21, 1981; however, the leak was in the underground portion of the Unit 1 RHR loop "A" Service Water line.

A similar Technical Specification change was requested to allow the use of the cross-tie valve while rerouting the piping above ground and abandoning the underground cracked piping.

TJkILUM2fECBhN1SB It is surmised that the pipe was pierced by a temporary construction support that was not removed during the backfill operation during original plant construction. This is further supported due to routing schemes of the RHR piping. The Unit 1 "A" loop and the Unit 2 "B" loops are the lower lying piping. To date, cracks have been identified in each loop. The proposed modification to the Unit 2 "B" loop and the completed modification on the Unit 1 "A" loap should ensure no further cracking of the piping due to this failure mechanism.

ERQERSED MODIFICATION As indicated previously, a similar modification is proposed for the Unit 2 "B" loop of RHR Service Water as was performed on the Unit 1 "A" loop of the RHR Service Water in 1982.

New piping for the Unit 2 "B" loop will be routed above ground.

Following installation of the new pipe, the "B" loop of the Unit 2 RHR Service Water System will be taken out of service and an end cap installed. The isolated piping will be abandoned in place. The new piping will be tied into the system. During the installation of the end cap and tie in of the new piping into the system, the 7 day Limiting Condition for Operation (LCO) timeclock will be in effect.

i SYSTEM OPERABILITY The-RHR Service Water system is operable. A surveillance was . j conducted on March 13, 1989 to demonstrate operability based on Technical  ;

Specification flow requirements and was acceptable. In order to monitor.the any possible further deterioration of the crack, a weekly surveillance will be conducted. Increased attention to the flow rates will be focused in order to identify degradation of the pipe. A Safety Evaluatlon has been conducted to ensure that the crack does not reduce any margin of safety _to continued operation.

BASIS _fDiLCHhtLGE Technical Specification LCO 3.5.B.1(a) requires two pumps and one heat exchanger for each unit's RHR Service Water loop. The system configuration proposed on the Unit 1 and 2 "B" loop will be such that two ,

pumps will be available to provide flow'to either unit's heat exchanger-(Figure 1). 'The "A" loop configuration on Units 1 and 2 remain unchanged.

The RHR Service Water system configuration required to provide adequate containment cooling capability-following a loss of coolant accident is described in the Safoty Analysis Report (SAR) and consists of one.RHR Service Water pump and one RHR pump. The same configuration of equipment is adequate on a non-accident unit to place and maintain the~ reactor in a cold shutdown ,

condition. The minimum combination of equipment described in the SAR is only j experienced during degraded plant conditions, i.e., loss of offsite power, loss of coolant accident on one unit and a failure of one diesel generator to start. The proposed change is therefore justified since the Unit 1 and 2 RHR "B" loop service water system cross-tied configuration has been previously analyzed as acceptable.

. TIME E RIDP_RAEEE The proposed Technical Specification will be in effect.from the approval date of the change to April 1, 1990. A similar modification was completed in.1982 and the duration to complete the modification design and ,

installation was approximately eight (8) months. Since 1981, Commonwealth Edison has revised the process for modification design requiring additional reviews and controls. The period of one year has been determined to be a realistic schedule based on these changes.

ADDlIIDtiALC0HIROLEEDUIRED In the event that the cross-tie of the "B" loop is required, Quad Cities is currently reviewing the followings i

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(a) The need to develop procedures to provide guidance to Operations personnel concerning the cross-tied configuration, altered surveillance requirements and outage report' requirements.

.(b')- The impact on Appendix R procedures. j i

(c) Administrative controls to ensure that operators are aware that a loss of the Unit 1 C" or "D" RHR Service Water pump will render one containment cooling loop inoperable.

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CDECM EIDH As indicated previously, the Unit 2 "B" loop of the RHR Service Water system remains operable. The basis for operability of the system has been j discussed with both Region III and NRR personnel with their concurrence. l

'l The implementation of this Technical Specification revision will be J required only if the pipe crack has deteriorated to the extent that the RHR Service Water pumps can no longer meet the required flowrate dictated by

- Technical Specification Surveillance Requirement 3.5.B.1.b or it has been determined that the isolation of the crack is prudent.

.i This Technical Specification is submitted for review in the event of further pipe degradation.

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4 AITACHMENT.4 l

EYALUATION OF SIGNIrlCART_ IIA'dARD1_CRHSIDERATION As stated in the " Description of Proposed Amendment Request", the proposed change involves the use of the unit cross tie for the "B" loop of the.

RHR Service Water system and the use of the Unit 1 "C" and "D" RHR Service Water pumps to supply either unit's "B" loop. These changes have been L reviewed by Commonwealth Edison and the changes do not present a Significant l - Hazards Consideration.

I HASIS FOR NO SIGNIFICANT BAZARDS CONSIDERAT1Qti f i

In accordance with the criteria of 10 CFR 50.92(c), Commonwealth )

Edison has reviewed the proposed snendment and has determined that it involves l no significant hazards consideration, i (1) The proposed change will not involve a significant increase in the 1 probability or consequences of an accident previously evaluated. l The RHR Service Water system is designed to mitigate the consequences of an accident by removing decay heat from the containment, therefore, the probability of occurrence of an accident is not increased by failure of any components in this system.

As described in Section 6, Amendments 16 and 17 to the Quad Cities j Safety Analysis Report (SAR), one RHR pump and one RHR Service Water pump provide adequate containment cooling following a loss of coolant accident. A similar combination of equipment is adequate on a non-accident unit to place and maintain the reactor in the cold shutdown condition. This minimum combination of equipment is only experienced in the degraded conditions of loss of offsite power, loss of coolant accident on one unit, and failure of one diesel generator to start. Since this modification does not reduce the minimum RHR Service Water system availability as described in the SAR, the consequences of an accident or malfunction of equipment important to safety have not increased. As will be discussed in detail later, redundancy of RHR Service Water pumps remains consistent with the original design.

(2) The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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This cross-tie modification shares two RHR Service Water pumps between Units 1 and 2 in the same manner that one of the emergency power supplies, the 1/2 Diesel Generator, is shared between the two j units. The cross-tie does not involve the addition of any active  !

components,' electrical interlocks, etc. The cross-tie piping and

, manual valve are existing safety related parts of the RHR Service Water system. The new pipe cap will. maintain the pressure boundary of the RHR Service Water-system and will be designed to equivalent- ,

standards to the originni piping. Similar use of a cross-tie on the J "A" loop of each unit's RHR Service Water system was performed in 1981.

(3)' The, proposed change does not involve a significant reduction in the '

margin of safety.

-i In the cross-tied condition,.the worst scenario involves and accident on Unit 1, loss of offsite power, and failure of the 1/2 Diesel j Generator. After core cooling is restored by the ECCS on Unit 1, the I required loads to provide RHR Service Water to both units and  !'

maintain core cooling on the accident unit are within the capability of the Unit 1 Diesel Generator.

These conditions result in the minimum operability of two RHR Service Water pumps, one per unit as analyzed in the SAR.

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A'T T A C H M'E N'T -5 p atooq'o.

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g a E' NUCLEAR REGULATORY COMMISSION WASHINGTON, D, C. 20555 SAFETY EVALUATION BY THE'0FFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 75 TO PROVISIONAL OPERATING LICENSE N0. DPR-29 AMENDMENT N0. 68 TO FACILITY OPERATING LICENSE N0 ' DPR-30 COMMONWEALTH EDISON COMPANY l-AND IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY j QUA0 CITIES STATION UNIT NOS.1 AND 2 DOCKET NOS. 50-254 AND 50-265 INTRODUCTION By letter dated December 3,1981 Commonwealth Edison Company (the licensee) I proposed a temporary change to Appendix A. Technical. Specifications, to i Facility Operating Licenses DPR-29 and DPR-30 for Quad Cities Units 1 and 2, respectively.

On November 21, 1981 a leak was discovered in the underground portion of the Unit 1 RHR loop "A" Service Water Line. Unit 1 was in an outage at the' . . l O- time and could not restart because of failure to meet a Technical Specification j

(T/S) requirement for two operable RHR containment cooling-loops. Ii1 order to meet the T/S requirement, the RHR Service Water Pumps "A" and "B" from Unit 2 were made available to Unit 1 by utilizing a cross-tie line.(see Figure 1). Unit 2 was in a refueling outage at the time and this' equipment I (RHR SW pumps A & B) was not required to be operable for Unit 2. However,  !

the T/S precludes any refueling work that has the potential- for. draining the vessel unless the low-pressure core cooling and containment cooling systems are operable.

EVALUATION The licensee proposes a temporary T/S modification that allows the RHR containment cooling loop "A" for Unit 2 to be defined as operable for Unit 2 while the cross-tie line is connected to Unit 1. The'T/S amendment will allow the use of the cross-tie until June 1,1982, at which time the i repair of the Unit 1 RHR loop "A" service water line will be completed.

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, 'The containment cooling mode of the RHR system for each unit consists of ,

two loops as shown in Figure 1. Each loop consists of one heat exchanger, l two RHR pumps, associated valves, piping, electrical equipment and instru-menta tion. The "B" loop on each unit contains two RHR service water pumps.

Normally the "A" loop on each unit also contains two RHR service water pumps. However, during the time interval from November 24, 1981 to June 1, 1982 the "A" loop on each unit may utilize the "A" and "B" RHR service

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water pumps from Unit 2. Service water from Unit 2.to Unit I will be u

delivered via the cross-tie line. Loss of one loop of the containment cooling mode of the RHR system leaves the remaining loop to perform the i containment cooling function. Either loop of the RHR system can satisfy the containment cooling function.

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As stated in the Quad Cities FSAR, only one RHR and one RHR service water pump are required to provide containment cooling following a loss-of-coolant accident (LOCA) on one unit. One RHR and one RHR service water pump are also adequate to place and maintain the other unit in the cold shutdown condition. Adequate containment cooling is therefore available in the event of a design basis LOCA concurrent with a loss of offsite power (LOOP) anc the worst single failure, if one RHR and one RHR service water pump are available. If the Unit 1/2 (swing) D/G or Unit 2 D/G is postulated as the single failure, along with a loss of Offsite Power, the modification results in no change to existing procedures or to the existing design basis. If the Unit 1 D/G is postulated to fail, the original design ,.

basis is met with the proposed changes since procedure changes and electrical modifications have been proposed to assure that adequate power is allocated ,

to the necessary equipnent during a Loss of Offsite Power.

A procedure will be implemented which will allow Buses 13-1 and 23-1 to be energized by D/G 1/2 at the same time (see Figure 2). This will permit one Unit 2 RHR service water pump and one Unit 1 RHR pump to be powered i by the D/G 1/2. This procedure would require that the ECCS pumps on  !

Bus 23-1 be pulled-to-lock (prevented from auto starting) prior to closing i two locally mounted control switches (as described below) which permit both buses to be energized by the 1/2 diesel .

n To permit 1/2 D/G to supply Bus 13-1 on Unit 1 and Bus 23-1 on Unit 2 at (V) the same time, a breaker position interlock in the closing circuit of either one of the two 1/2 DG output breakers must be bypassed when the I redundant breaker is already closed. This action is accomplished in either ,

1/2 DG output breaker by means of a control switch which is physically and '

electrically independent of its redundant counterpart. Once the breaker position interlock is bypassed in the 1/2 DG output breaker that is open, ,

the operator will then be able to close this output breaker from the control room. {'

To minimize the probability of operator errors that will leave the bypass control switches in the wrong position, after their function was accomplished, ,

the licensee was advised that we will require that the bypass condition in each breaker closing circuit be automatically removed when the associated 1/2 DG output breaker is opened. The licensee has elected to implement this requirement by providing bypass control switches that are spring-return-to-open. This in essence, automatically removes the bypass as soon as the operator releases the switch from the bypass position. We find this to be an acceptable way to implement our requirement in this regard. The licensee has also stated that the bypass control switches will be of the same type as those previously qualified and presently used in existing safety related circuits in the plant.

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0 l The licensee stated that the modification will be fully tested to assure that, the 1/2 diesel can energize Bus 23-1 when it is already energizing Bus 13-1, or energize Bus 13-1 when it is already energizing Bus 23-1. It will also ' ,

be demonstrated that the closing circuits will revert to the original  :

condition when the bypasses are removed.

Prior to this modification implementation, the licensee has committed to a write procedures and train operators in the use of this new installation. i 1

We find this temporary modification acceptable because the licensee states and we concur based upon information elicited from the licensee that:

(1) there is ample time, and there will be appropriate procedures and operator training to manually actuate the local switches to initiate the containment cooling function after an event; (2) the RHR containment q cooling mode was always designed as a manual action; (3) no originally L designed automatic protective features will be adversely affected; and h (4) there will be no compromises to core cooling by this temporary [

modi fication. L Environmental Considerations .

We have detemined that these amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that these amendments involve an action which is insignificant from the standpoint of environmental impact, and pursuant to 10 CFR Section 51.5(d)(4) that an . environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.

Conclusion h

We have concluded, based on the considerations discussed above, that: (.1)because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a signifi-cant decrease in a safety margin, the amendments do not involve a significant c hazards consideration, (2} there is reasonable assurance that the health and  :

safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Coranission's regulations and the issuance of these amendments will not be inimical to the connon defense and security or to the health and safety of the public.

Dated: December 18, 1981 q

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