ML20215N863

Proposed Tech Specs,Changing LPCI Flow Test Requirements from Current Three Pump Test Demonstrating 14,500 Gpm Flow Rate to Two Pump Test Demonstrating 9,000 Gpm Flow Rate. NSHC Evaluation Encl

ML20215N863
Person / Time
Site:	Quad Cities
Issue date:	10/22/1986
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20215N861	List: ML20215N860 ML20215N863
References
NUDOCS 8611100006
Download: ML20215N863 (15)

Text

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PROPOSED CHANGES TO OUAD CITIES UNIT 1 TECHNICAL SPECIFICATIONS i

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1 2216K 8611100006 861022 PDR ADOCK 05000254 P. PDR

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QUAD-CITIES DPR-29

e. Core spray header a p instrumentation check Once/ day calibrate Once/3 months test Once/3 months

f. Logic system Each functional refueling test outage

2. From and after the date that one 2. When it is determined that one of the core spray subsystems is core spray subsystem is inoper-made or found to be inoperable able, the operable core spray for any reason, continued reac- subsystem, the LPCI mode of the tor operation is permissible RHR system, and the diesel gen-only during the succeeding 7 erators required for operation days unless such subsystem is of such components if no exter-sooner made operable, provided nal source of power were avall-that during such 7 days all~ac- able shall be demonstrated to be tive components of the other operable immediately. The oper-core spray subsystems and the able core spray subsystem shall LPCI mode of the RHR system and be demonstrated to be operable the diesel generators required daily thereafter.

for operation of such components if no external source of power were available shall be operable.

3. The LPCI mode of the RHR system 3. LPCI mode of the RHR system shall be operable whenever testing shall be as specified in irradiated fuel is in the Specifications 4.5.A.I.a. b, c, reactor vessel and prior to d, and f, except that each LPCI reactor startup from a cold division (two RHR pumps per condition. division) shall deliver at least 9000 gpm against a system head corresponding to a reactor ves-sel pressure of 20 psig, with a minimum flow valve open. '

4. From and after the date that one of the RHR pumps is made or 4. When it is determined that one found to be inoperable for any of the RHR pumps is inoperable, reason, continued reactor opera- the remaining active components tion is permissible only during of the LPCI mode of the RHR, the succeeding 30 days unless containment cooling mode of the such pump is sooner made oper- RHR, both core spray subsystems, able, provided that during such and the diesel generators re-30 days the remaining active quired for operation of such components of the LPCI mode of components if no external source the RHR, containment cooling of power were available shall be demonstrated to be operable immediately and the operable RHR pumps daily thereafter.

0658H 3.5/4.5-2 Amendment No.

QUAD-CITIES DPR-29 mode of the RHR, all active components of both core spray subsystems, and the diesel generators required for operation of such components if no external source of power were available shall be operable.

5. From and after the date that the 5. When it is determined that the LPCI mode of the RHR system is LPCI mode of the RHR system is made or found to be inoperable inoperable, both core spray sub-for any reason, systems.-the

.0658H 3.5/4.5-2a Amendment No.

QUAD-CITIES DPR-29 3.5 LIMITING CONDITION FOR OPERATION BASES A. Core Spray and LPCI Mode of the RHR System This specification assures that adequate emergency cooling capability is available whenever irradiated fuel is in the reactor vessel.

Based on the loss-of-coolant analytical methods described in General Electric Topical Report NED0-20566 and the specific analysis in Reference-1, core cooling systems provide sufficient cooling to the core to dissipate the energy associated with the loss-of-coolant accident, to limit calculated fuel cladding temperature to less than 22000F, to assure that core geometry remains intact, to limit cladding metal-water reaction to less than 17., and to limit the calculated local metalwater reaction to less than 17%.

The limiting conditions of operation in Specifications 3.5.A.1 through 3.5.A.6 specify the combinations of operable subsystems to assure the availability of 're minimum cooling systems noted above. Under these limiting Conditions of operation, increased surveillance testing of the remaining ECCS systems provides assurance that adequate cooling of the core will be provided during a loss-of-coolant accident.

Core spray distribution has been shown, in full-scale tests of systems similar in design to that of Quad-Cities 1 and 2, to exceed the minimum requirements by at least 257.. In addition, cooling effectiveness has been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irradiated fuel. The accident analysis is additional conservative in that no credit is taken for spray cooling of the reactor core before the internal pressure has fallen to 90 psig.

The LPCI mode of the RHR system is designed to provide emergency cooling to the core by flooding in the event of a loss-of-coolant accident. This system functions in combination with the core spray system to prevent excessive fuel cladding temperature. The LPCI mode of the RHR systems in combination with the core spray system to prevent excessive fuel cladding temperature. The LPCI mode of the RHR system in combination with the core spray subsystem provides adequate cooling for break areas of approximately 0.2 ftZ up to and including 4.18 ft 2, the latter being the double-ended recirculation line break with the equalizer line between the recirculation loops closed without assistance from the high-pressure emergency core cooling subsystems.

The allcwable 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 found to be less than 00658H 3.5/4.5-11 Amendment No.

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QUAD-CfTIES DPR-29 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 llmit excessive cladding temperature must 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 1 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 establis'hed with due regard to judgment.

Should one core spray subsystem become inoperable, the remaining core.

spray subsystem and the entire LCPI mode of the RHR system are available should the need for core cooling arise. To assure that the remaining core spray, the LPCI mode of the RHR system, and the diesel generators are available, they are demonstrated to be operable immediately. This demonstration includes a manual initiation of the pumps and associated valves and diesel generators. Based on judgements of the reliability of the remaining systems, i.e., the core spray and LCPI, a 7-day repair period was obtained.

1 0658H 3.5/4.5-lla Amendment No.

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ATTACHMENT 2 PROPOSED CHANGES TO OUAD CITIES UNIT 2 TECHNICAL SPECIFICATIONS I

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QUAD-CITIES DPR-30

e. Core spray header a p instrumentation check Once/ day calibrate Once/3 months test Once/3 months

f. Logic system Each functional refueling test outage

2. From and after the date that one 2. When it is determined that one of the core spray subsystems is core spray subsystem is inoper-made or found to be inoperable able, the operable core spray for any reason, continued reat- subsystem, the LPCI mode of the tor operation is permissible RHR system, and the diesel gen-only during the succeeding 7 erators required for operation days unless such subsystem is of such components if no exter-sooner made operable, provided nal source of power were avail-that during such 7 days all ac- able shall be demonstrated to be tive components of the other operable immediately. The oper-core spray subsystems and the able core spray subsystem shall LPCI mode of the RHR system and be demonstrated to be operable the diesel generators required daily thereafter.

for operation of such components if no external source of power were available shall be operable.

3. The LPCI modo of the RHR system 3. LPCI mode of the RHR system shall be operable whenever testing shall be as specified in irradiated fuel is in the Specifications 4.5.A.1.a. b, c, reactor vessel and prior to d, and f, except that each LPCI reactor startup from a cold division (two RHR pumps per condition. division) shall deliver at least 9000 gpm against a system head corresponding'to a reactor ves-sel pressure of 20 psig, with a minimum flow valve open.

4. From and after the date that one of the RHR pumps is made or 4. When it is determined that one found to be inoperable for any of the RHR pumps is inoperable, reason, continued reactor opera- the remaining active components tion is permissible only during of the LPCI mode of the RHR, the_ succeeding 30 days unless containment cooling mode of the such pump is sooner made oper- RHR, both core spray subsystems, able, provided that during such and the diesel generators re-30 days the remaining active quired for operation of such components of the LPCI mode of components if no external source the RHR, containment cooling of power were available shall be demonstrated to be operable immediately and the operable RHR pumps daily thereafter.

3.5/4.5-2 Amendment No.

0658H

,* QUAD-CZTIES OPR-30 mode of the RHR, all active components of both core spray subsystems, and the diesel generators required for operation of such components if no external source of power were available shall be operable.

5. From and after the date that the 5. When it is determined that the LPCI mode of'the RHR system is LPCI mode of the RHR system-is made or found to be inoperable inoperable, both core spray sub-for any. reason, systems, the 3.5/4.5-2a Amendment No.

0658H

ATTACIWWNT ).

DESCRIPTION OF PROPOSED CHANGE

'I. BACKGROUND As part of Commonwealth Edison's long term solution to the single failure concern for BWRs with low pressure coolant injection (LPCI) loop selection logic addressed in IRB 86-01, it is proposed to modify the LPCI/RHR pump minimum flow valve control logic. The modification i

will remove the interlock between Divison I and Division II.

, Currently the 'A' and 'B' minimum flow valves ( one minimum flow valve for each set of two RHR/LPCI pumps) are controlled by either flow sensor 'A' or 'B'. This allows failure of the A or B flow sensor to cause a false signal to both'A and B minimum flow valves, which could result in the RHR pumps running " dead headed" and potential failure of both sets of pumps. The proposed design modification will change the

, control logic such that minimum flow valve A is controlled only by j- flow sensor A and minimum flow valve B is controlled only by flow sensor B. This modification will eliminate the single failure concern addressed in IRB 86-01 and is consistent with the General Electric (GB) recommendations in SIL No. 444.

After incorporation of the modification, during LPCI operation the

flow sensor in the selected loop will signal the corresponding minimum

' flow valve to close. However, the other division's minimum flow valve

-will remain open reducing the rate of flow to the core by an amount equivalent to the flow through the minimum flow line.

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l In order to assess the impact of the flow loss through the minimum flow line, Quad Cities station performed several tests to demonstrate the amount of flow that the RHR/LPCI pumps are capable of in the LPCI mode, for a corresponding reactor pressure of 20 psi, with the minimum flow valves open. This test indicated that,'after the modification is

[. complete, the pumps will not be able to meet the two pump 9667 gpm j flow rate assumed in the Appendix K Loss-of-Coolant Accident analysis I

(LOCA).

II. BASIS As a result of the planned LPCI minimum flow valve control logic t modification and the tests performed for that modification, Commonwealth Edison proposes to revise the Quad Cities Unit 1 and 2 Technical Specification Surveillance Requirement 4.5.A.3 and the corresponding bases. The proposed revision requires that a two pump RHR/LPCI test at a flowrate of 9000 gpm be performed every three months, rather than the current three pump 14,500 gpm test. A two pump test is proposed because it more accurately depicts the assumptions and requirements of the Appendix K LOCA analysis, which used more restrictive criteria than the original analysis.

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. i 4g In the current Appendix K LOCA analysis the most limiting event is t.he s ,

hypothetical, doubled-ended recirculation suction line break with an. (' ,'

assumed failure of the LPCI injection valve. This' failure assumes no credit for the LPCI pumps; therefore, a change in the LPCI flow has no effect on this event and the core cooling capability is not reduced for the most limiting event. ,

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i, s The second most 1Leiting break and single failure combination, which takes credit for LPCI cooling, is the DBA recirculation sudtlon sline _,,

break.with a diesel generator failure. jntis scenario requires one low

• pressure core spray pump and two LPCI pumps for core cool,ing. GE has analyzed this event assuming the proposed'9000 gpm flowrate'for two LPCI pumps rather than the 9667 gpm flowrate assumed in the $nitial Appendix K analysis. The results of this analysist(Attachment 4) 1

indicate.that with the modification the peak cladding temperature

i. (PCT) increases by 32*F to 1793'F. This temperature is well below the ,

Quad Cities limiting break PCT and the 2200*F limit set by the i Appendix K requirements.

The other modes of RHR were also reviewed against the proposed change. These other modes of the RHR system are unaffected by the modification since all other RHR modes are manually actuated and have relatively long required operator response times. Therefore, there is adequate time for the operator to diagnose and correct the situation if the valves are found open.

III. CONCLUSIONS Commonwealth Edison has reviewed the proposed'ch'nge a to the Technical 4

Specification and the supporting GB analyses, which shows that the existing LOCA design basis is unchanged, and concludes that the analysis is acceptable to support the proposed revision and that no unreviewed safety questions exist.

2216K

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ATTACHMENT 4 GENERAL ELECTRIC

SUMMARY

REPORT FOR REDUCED LPCI FLOW 2216K

SUMARY OF QUAD CITIES I/II REDUCED LPCI Fim IVALUATION August 1986 A planned plant modification at had Cities will permit the Residual Heat Removal (RHR) pu=p mini:m.:m finv hypass lines to be open during a Loss-of

-Coolant Accident (LOCA). This has the potential for reducing the Low Pressure Coolant Injection (LPCI) mode flowrate to the vessel to a value less than that assumed in the current licensing basis. Therefore, the impact of the reduced LPC1 flowrate on the limiting pipe breaks and single failures must be investigated.

For the most limiting accident event, the core cooling capability is not reduced. The limiting LOCA is a hypothetical, double ended recirculation suction line break with an assumed failure of .the LPCI injection valve.

For this event, no credit is assumed for LPCI cooling; core cooling is provided by the two core spray pumps. An evaluation of the Ifmiting breaks and single failures that show sensitivity to LPCI flowrate shows that the existing LOCA design basis is unchanged. This evaluation assumsd a Technical Specification minimum flow of 9000 gpm for two pumps operating in one loop.

The limiting break and single f ailure combination that includes LPCI cooling is the DBA recirculation suction line break with a diesel generator (DG) failure. For this event the Peak Cladding Temperature (PCT) increases by 32 deg-F with the modification but is still well below the limiting break PCT of 2200 deg-F (see Table 1). Additionally, the other modes of the RHR system are unaffected by the modification since all other RHR modes are manually actuated and have relatively long required operator response times. Therefore there is adequate time for the operator to diagnose and correct the situation if the valves are found open.

l. 1/&

E.H. Hof fmaMi, Engineer Application Engineering Services Verified by: Approved by:

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, l14 (Dw) G.L. Sozzi, Manager '

P.T. Tran, ingineer Application Engineering Services Application Engineering Services DRF A00-2423 i

TABLE 1 RESULTS OF PROPOSED MODIFICATION ON QUAD CITIES CORE COOLING ANALYSIS PEAK CLADDING TEMPERATURE (*F)

ATTER SYSTT.riS BEFORE SUCTION LINE SINGLE MODIFICATION FAILURE AVAILABLE MODIFICATION BREAX SIZF_

2CS 2200 2200 4.18 ft2 (DBA) LPCIIV 1761 1793 4.18 ft (DBA) DG ICS+2LPCI 2

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ATTACMENT 5 SIGNIFICANT HAZARDS CONSIDERATION EVALUATION DESCRIPTION OF AMENDMENT REQUEST ,

Commonwealth Edison has requested an amendiisent to the Quad Cities Units 1 and 2 Technical Specifications to modify the Low Pressure Coolant Injection (LPCI) pump flow test requirements from the current three pump test demonstrating 14,500 gpm to a two pump test demonstrating 9000 gps.

This change is required to support a modification to the LPCI pump minimum flow valve control logic to resolve a single failure concern identified in I.F.. Bulletin 86-01.

DISCUSSION The existing LPCI loop selection logic is such that failure of either the "A" loop or "B" loop flow sensor would close both the "A" and "B" minimum flow valves. In response to I.E. Bulletin 86-01, this logic is being acdified so that the "A" valve is controlled only by the "A" flow sensor.and the'"B" valve by the "B" sensor. With this configuration, during LPCI injection through the loop crosstie the minimum flow valve on the operating LPCI loop would remain open, thereby reducing the LPCI loop flow (consisting of two LPCI pumps) by the amount of recirculating flow through the minimum flow valve.

4 The original design of the LPCI system required three pumps with a conbined flow capacity of 14,500 gpm. In 1974, 10 CFR 50.46/ Appendix K was implemented requiring re-analysis of the Design Basis Loss of Coolant Accident (DBLOCA) to demonstrate compliance with the revised peak clad temperature limit (2200*F) and single failure criterion. This is the current Emergency Core Cooling System licensing basis. In the current Appendix K LOCA analysis, the most limiting event-is the hypothetical, double-ended recirculation suction line break with an assumed failure of the LPCI injection valve. .This scenario assumes no credit for the LPCI pumps; therefore, the proposed Technical Specification change in the LPCI flow has no effect on this event and the core cooling capability is not reduced for the most limiting event.

The second most limiting break and single failure combination, which takes credit for LPCI cooling, is the DBA recirculation suction line break with a diesel generator failure. This scenario requires one low pressure core spray pump and two LPCI pumps for core cooling. General Electric has analyzed this event assuming the proposed 9000 gpm flowrate for two LPCI pumps rather than the 9667 gpm flowrate assumed in the initial Appendix K analysis. The results of this analysis (Attachment 4) indicate that with the revised flowrate the peak cladding temperature (PCT) increases by 32*F to 1793*F. This temperature is well below the Quad Cities limiting break PCT and the 2200*F limit set by 10 CFR 50.46.

l,

l tsA313 tw NU SIGNit'1 CANT HAZANDS CONSIDt1NAT10N DtfrEMMINATION c e alth Edison has evaluated the proposed Technical Specifica-tion amendment and determined that it does not represent a significant hazards consideration. Based on the criteria for defining a significant hazards' consideration established in 10 CPR 50.92(c), operation of Quad Cities in accordance with the proposed amendment will not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated because the proposed LPCI flowrate does not affect the most limiting design basis LOCA and for the event that is affected, the slightly increased peak cladding temperature is well below the limiting event peak cladding t emperature and the limit set by the NRC. The probability of a LOCA is unaffected since the proposed change affects an accident

, mitigating system and has no relationship to the failures or events necessary to initiate an accident.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed change in flow rate requirements does not affect any systems or equipment whose failure or malfunction during operation could lead to the initation of a reactor transient or accident.

3. Involve a significant reduction in the margin of safety since the analysis supporting this Technical Specification change showed that existing margins to safety are preserved because the existing design basis LOCA is unchanged. The change in peak clad temperature for the affected non-limiting event is insignificant in comparison to the available margin.

3 Based on the previous discussion Commonwealth concludes that the proposed Technical Specification changes do not represent a significant hazards consideration.

2216K J

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