ML20065F725

PWR Main Steam Line Break W/Continued Feedwater Addition (B-69),Oconee Nuclear Station Units 1,2 & 3, Technical Evaluation Rept

ML20065F725
Person / Time
Site:	Oconee
Issue date:	09/28/1982
From:	Herrick R, Vosbury F FRANKLIN INSTITUTE
To:	Peter Hearn NRC
Shared Package
ML15238A677	List: ML20065F725
References
CON-NRC-03-81-130, CON-NRC-3-81-130 IEB-80-04, IEB-80-4, TAC-46847, TAC-46848, TAC-46849, TER-C5506-134, NUDOCS 8210010400
Download: ML20065F725 (23)

Text

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. t TECHNICAL EVALUATION REPORT PWR MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION G-69)

DUKE POWER COMPANY I OCONEE NUCLEAR STATION UNITS 1, 2, AND 3 NRC DOCKET NO. 50-269, 50-270, 50-287 FRC PROJECT C51106 NRC TAC NO. 46847, 46848, 46849 FRC ASSIGNMENT 5 NRC CONTRACT NO. NRC-03-81-130 FRC TASK 134 I

, Preparedby Franklin Research Center Author: F. W. Vosbury l 20th and Race Street i Philadelphia, PA 19103 FRC Group Leader: R. C. Herrick {

Prepared.for Nuclear Regulatory Commission Washington, D.C. 20555 Leed NRC Engineer: P. Hearn l

September 28, 1982 I

This report was 4 ' pared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability.or responsibility for any third party's use, or the results of such use, of any information, appa-ratus, product or process disclosed in this report, or represents that its use by such third party would not ir5fritt a privately owned rignts.

Prepared by: Reviewed by: Approved by:

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Date: 7 -M-b' Date: ff//E Date: 'l - 2 &-? 2.

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CONTENTS Title Page Section 1 INTRODUCTION. . . . . . . . . . . . . . 1 1.1 Purpose of Review . . . . . . . . . . . 1 1.2 Generic Background . .. . . . . . . . . . 1 1.3 Plant-Specific Background . . . . . . . . . 3 ACCEPTANCE CRITERIA . . . . . . 4 2 . . . . . .

TECHNICAL EVALUATION. . . . . . . . . . 8 3 . .

3.1 Review of Containment Pressure Response Analysis . . . 8 3.2 Review of Reactivity Increase Analysis . . . . . . 13 3.3 Review of Corrective Actions . . . . . . . . 16 4 COtCLUSIONS . . . . . . . . . . . . . .. 10 5 REFERENCES . . . . . . . . . . . . . . 19 l

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. f TER-C5506-13 4 FORBf0RD This Technical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. Nuclear Regulatory Comunission (Office of Nuclear Reactor Regulation, Division of Operating Reactors) for t % nical assistance in support of NRC operating reactor licensing actions. The technical evaluation was conducted in accordance with criteria established by the NRC.

Mr. F. W. Vosbury contributed to the technical preparation of this report through a subcontract with WESTEC Services, Inc.

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1. INTRODUCTION 1.1 PURPOSE OF REVIEN

. * .4 Technical Evaluation Report (TER) documents an independent review of Duke hwrer Company's (DPC) response to the Itaclear Regulatory Commission's (NRC) IE Bulletin 80-04, " Analysis of a Pressurized Water Reactor Main Steam Line Break with continued Feedwater Addition" [1], as it pertains to the j.

Oconee Nuclear Station Units 1, 2, and 3. This evaluation was performed with the following objectives:

o to assess the conformance of DPC's main steam line break (MSLB) analyses with the requirements of IE Bulletin 80-04 o to assess DPC's proposed interim and long-range corrective action plans and schedules, if needed, as a result of the MSLB analyses. '

l.2 GENERIC BACKGROUND In the summer of 1979, a pressurized water reactor (PWR) licensee .

submitted a report to the NRC that identified a deficiency in the plant's original analysis of the containment pressurization resulting from a MSLB. A reanalysis of the containment pressure response following a MSLB was performed, and it was determined that, if the auxiliary feedwater (APW) system continued to supply feedwater at runout conditions to the steam generator that had experienced the steam line break, containment design pressure would be exceeded in approximately 10 minutes. The long-term blowdown of the water supplied by the APW system had not been considered in the earlier analysis.

On October 1,1979, the foregoing information was provided to all holders I of operating licenses and constructicn permits as IE Information Notice 79-24

[2]. Another facility performed an accident analysis review pursuant to l receipt of the information in the notice and discovered that, with offsite l electrical power available, the condensate pumps would feed the affected steam h generator at an excessive rate. This excessive feed was not previously considered in the plant's analysis of a MSLB accident. j 1

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t TER-C5506-134 A third licensee informed the NRC of an errce in the MSIB analysis for their plant. During a review of the MSIa analysis, for zero or low power at the end of core life, the licensee identified an incorrect postulation that the startup feedwater control valves would remain positioned "as is" during i the transient. In reality, the startup feedwater control valves will ramp to 804 full open due to an override signal resulting from the low steam generator pressure reactor trip signal. ananalysis of the events showed that opening of the startup valve and associated high feedwater addition to the affected steam generator would cause a rapid reactor cooldown and resultant reactor return-to-power response, a condition which is outside the plant design basis.

i Because of these deficiencies identified in original MSIa accident analyses, the NBC issued IE Bulletin 80-04 on February 8,1980. This bulletin required all PWRs with operating licenses and certain near-tern PWR operating license applicants to perform the following:

"l. . Review the containment pressure response analysis to determine if the potential for containment overpressure for a main steam line break *

$ inside containment included the impact of runout flow from the e auxiliary feedwater system and the impact of other energy sources, such as continuation of feedwater or condensate flow. In your review, consider your ability to detect and isolate the damaged steam generator from these sources and the ability of the pumps to remain operable after extended operation at runout flow.

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2. Review your analysis of the reactivity increase which results from a j main steam line break inside or outside containment. This review i should consider the reactor cooldown rate and the potential for the j reactor to return to power with the most reactive control rod in the J fully withdrawn position. If your previous analysis did not consider j all potential water sources (such as those listed in 1 above) and if the reactivity increase is greater than previous analysis indicated, the report of this review should includes

a. The boundary conditions for the analysis, e.g., the end of life shutdown margin, the moderator temperature coefficient, power level and the net effect of the associated steam generator water inventory on the reactor system cooling, etc.,

b. me most restrictive single active failure in the safety injection system and the effect of that failure on delaying the delivery of high concentration boric acid solution to be reactor coolant system, i

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c. Se effect of extended water supply to the affected steam generator on the core criticality and return to power, i l

d. The hot channel factors corresponding to the most reactive rod in  !

the fully withdrawn position at the end of life, and the Minimum Departure from t&acleate Boiling Ratio (MDNBR) values for the i analyzed transient. -

3. If the potential for containment overpressure exists or the reactor return-to-power response worsens, provide a proposed corrective action and a schedule for ccepletion of the corrective action. If the unit is operating, provide a description of any' interim action that will be taken until the proposed corrective action is completed."

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1.3 PIJNT-SPECIFIC BACKGROUND i

Duke Power Company responded to IE Bulletin 80-04 in a letter to the NBC dated May 7, 1980 (3] and provided additional information for this review in a letter dated July 23, 1982 (4]. The information in References 3 and 4 has been evaluated along with pertinent information from the Oconee Nuclear Station Final Safety Analysis Report (FSAR) (S] to determine the adequacy of the Licensee's compliance with IE Bulletin 80-04. .

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2. ACCEPTANCE CRITERIA The following criteria against which the Licensee's MSLB response wu evaluated were provided by the NRC (6]

1. PWR licensees' responses to IE Bulletin 80-04 shall include the '

following information related to their analysis of containment pressure and core reactivity response to a MSLB within or outside containment:

a. A discussion of the continuation of flow to the affected steam

• generator, including the impact of runout flow from the ArW system and the impact of other energy sources, such as continuation of feedwater or condensate flow. APW system runout flow should be determined from the manufacturer's pump curves at no backpressure, unless the system contains reliable anti-runout provisions or a more representative backpressure has been conservatively calculated. If a licensee assumes credit for anti-runout provisions, then justification and/or documentation used to determine that the provisions are reliable should be provided. Examples of devices for which provisions are reliable are anti-runout devices that use active ccaponents (e.g., automatically throttled valves) which meet the requirements of IEEE Std 279-1971 (7] and passive devices (e.g. ,

flow orifices or cavitating venturis) .

b. A determination of potential containment overpressure as a result of the impact of runcut flow from the APW system or the impact of other energy sources such as continuation of feedwater or condensate flow. Where a revised analysis is submitted or where reference is made to the existing FSAR analysis, the analysis must show that runout APW flow was included and that design containment pressure was not exceeded.

c. A discussion of the ability to detect and isolate the damaged steam i generator from continued feedwater addition during the MSIS accident.

Operator action to isolate AFW flow to the affected steam generator -

within the first 30 minutes of the start of the MSIB should be justified. If operator action is to be completed'within the first 10 minutes, then the justification should address the indication available to the operator and the actions required. Where operator action is required to prevent exceeding a design value, i.e.,

containment design pressure or specified acceptable fuel design ,

limits, then the discussion should include the calculated time when tLo design value would be exceeded if no operator action were ,

assumed. Where operator actions are to be performed between 10 and  ;

30 mintues after the start of the MSLB, the justification should i

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, address the indications available to the operator and the operator actions required, noting that for the first 30 minutes, all actions should be performed from the control room.

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d. Where all water sources were not considered in the previous analysis, l an indication should be provided of the core reactivity change which j results from the inclusion of additional water sources. A submittal i which does not determine the magnitude of reactivity change from an  !

original analysis is not responsive to the requirements of IE Bulletin 80-04.

2. If containment overpressure or a worsening of the reactor-return-to-power with a violation of the specified acceptable fuel design limits described in Section 4.2 of the -Standard Review Plan [8] (i.e.,

increase in core reactivity) can occur by the licensee's analysis, the licensee shall provide the following additional informations

a. De proposed corrective actions to prevent containment overpressure or the violation of fuel design limits and the schedule for their completion. ,

b. De interim actions that will be taken until the proposed corrective action is completed, if the unit is operating.

3. Se acceptable input assumptions used in the licensee's analysis of the core reactivity changes during a MSLB.are given in Section 15.1.5 of the Standard Review Plan [9]. S e following specific assumptions should be used unless the analysis shows that a different assumption is more limiting:

Assumption II.3.b. : Analysis should be performed to determine the most conservative assumption with respect to a loss of electrical power. A reactivity analysis should be conducted for a normal power situation as well as a loss of offsite power scenario, unless the licensee has previously conducted a sensitivity analysis which demonstrates that a particular assumption is more conservative.

Assumption II.3.d.: 2e most restrictive single active failure in the safety injection system which has the effect of delaying the delivery of high concentration boric acid solution to the reactor coolant system, or any other single active failure affecting the plant response, should be considered.

Assumption II.3.g.: Se initial core flow should be chosen such that the post MSLB shutdown margin is minimized (i.e., maximum initial core ficw) .

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TER-C5506-134 The acceptable computer codes for the licensee's analysis of core reactivity changes are, by nuclear steam supply system (NSSS) vendor, the following: CESEC (Combustion Engineering), IDFTRAM (Westing- ..

house) , and TRAP (Babcock & Wilcox) . Other computer codes may be used, provided that these codes have previously been reviewed and found to be acceptable by the NRC staff. If a computer code is used which has not been reviewed, the licensee must describe the method employed to verify the code results in sufficient detail to permit

! the code to be reviewed for acceptability.

4. If the AFW pumps can be damaged by extended operation at runout flow, the licensee's action to preclude damage should be reviewed for technical merit. Any active features should satisfy the requirements of IEEE Std 279-1971. Where no corrective action has been proposed, this should be indicated to the NRC for further action and resolution.

! 5. Modifications to the electrical instrumentation and controls needed to detect and initiate isolation of the affected steam generator and j

feedwater sources in order to prevent containment overpressure and/or unacceptable core reactivity increases must satisfy safety-grade I requirements. Instrumentation that the operator relies upon to follow the accident and to determine isolation of the affected steam generator and feedwater sources should conform to the criteria contained in ANS/ ANSI-4.5-1980, " Criteria for Accident Monitoring Functions in Light-Water-cooled Reactors" (10), and the regulatory positions in Regulatory Guide 1.97, Rev. 2, " Instrumentation for j Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs '

Conditions During and Following an Accident" (11].

6. AFW system status should be reviewed to ensure that system heat  !

comoval capacity does not decrease below the minimum required level ,

as a result of isolation of the affected steam generator and also  !

that recent changes have not been made in the system Which adversely affect vital assumptions of the containment pressure and core reactivity response analyses.

7. The safety-grade requirements (redundancy, seismic and environmental qualifications, etc.) of the equipment that isolates the main feedwater (MFW) and AFW systems from the affected steam generator should be specified. The modifications of equipment that are relied upon to isolate the MFW and APW systems from the affected steam generator should satisfy the following criteria to be considered safety-grades o Redundancy and power source requirements: The isolation valves should be designed to accommodate a single failure. A failure-modes-and-effects analysis should demonstrate that the system is capable of withstanding a single failure without loss of function.

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. t TER-C5506-13 4 The single failure analysis should be conducted in accordance with the appropriate rules of application of ANS-51.7/N658-1976,

" Single Failure Criteria for PWR Fluid Systems" (12].

o Seismic requirements: The isolation valves should be designed to Category I as recommended in Regulatory Guide 1.26 [13] .

o Environmental qualification: The isolation valves should satisfy the requirements of NUREG-0588, Rev.1, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment" [14].

o Quality standards: The isolation valves should satisfy Group B quality standards as recommended in Regulatory Guide 1.26 or similar quality standards from the plant's licensing bases.

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3. TECHNICAL EVALUATION The scope of work included the following:

1. Review the Licensee's response to IE Bulletin 80-04 against the acceptance criteria.

2. a. Evaluate the Licensee's MSLB analyses for the potential of overpressurizing the containment and with respect to the core reactivity increase due to the effect of continued feedwater flow.

b. Evaluate the Licensee's proposed corrective actions and schedule for implementation if the findings of Task 2a indicate that a potential exists for overpressurizing the containment or worsening the reactor return-to-power in the event of a MSLB accident.

3. Prepare a TER for each plant based on the evaluation of the information presented for Tasks 1 and 2 above.

This report constitutes a TER in satisfaction of Task 3. Sections 3.1 through 3.3 of this report state _the requirements of IE Bulletin 80-04 by subsection, susmarize the Licensee's statements and conclusions regarding these requirements, and present a discussion of the Licensee's evaluation followed by conclusions and recossendations.

3.1 REVIEW OF 0:)NTAI!90DIT PRESSURE RESPONSE ANALYSIS The requirement from IE Bulletin 80-04, Item 1, is as follows:

" Review the containment pressure response analysis to determine if the potential for containment overpressure for a main steam line break inside containment included the impact of runout flow from the auxiliary feedwater system and the impact of other energy sources, such as continuation of feedwater or condensate flow. In your review, consider i your ability to detect and isolate the damagsd steam generator from these I sources and the ability of the pumps to remain operable after extended I operation at runout flow."

3.1.1 Summary of Licensee statements and conclusions In regard to the review of the containment pressure response analysis, the

, Licensee stated (3):

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. t TER-C5506-134 "The Oconee FSAR containment pressure response analysis for a postulated main steam line break inside containment considered two cases. The pertinent assumptions and system responses in each case are as follows:

Case 1 l '

Following the main steam line break, feedwater was assumed to remain at 1004 until the reactor trip occurred. Subsequently, the integrated control system (ICS) was assumed to close the main and startup control valves. The operator was assumed to take manual control of the feedwater to the affected OTSG [once-through steam generator] and ensure that the main and startup feedwater control valves remained closed. The unaffected OTSG was controlled at the minimum level (two feet) removing the decay heat generated in the core. The containment pressure response analysis calculated a 13 psi rise in the contaiwnt pressure, well below the design pressure of 59 psi."

Case 2 The second case was summarized in Reference 4 and is presented below:

" Assuming no operator action, feedwater will be delivered to the affected steam generator by the MrWS [ main feedwater system] following a main steam line break to control steam generator level at the setpoint. More energy will be delivered to the Reactor Building with the MFWS rather than the EFWS [ emergency feedwater system] in operation, due to the higher fluid enthalpy and higher flow capacity. As the Reactor Building pressure increases, the Reactor Building Spray System (RBSS) and the Reactor Building Cooling System (RBCS) will actuate and begin to remove energy from the building. These two systems are described in the Oconee FSAR. The feedwater delivered to the affected steam generator will continue to boil off and cool down the primary system. The increase in Reactor Building pressure causes the saturation temperature to increase,

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thereby decreasing the primary to secondary temperature difference across the steam generator tubes. This causes the heat transfer from the primary system to become limited to the heat being added to the primary system, which is the reactor decay heat. The Reactor Building pressure will continue to increase until the energy addition to the building is ,

j less than the energy removal by the RBSS and the RBCS. .

The results of the FSAR analysis show that at 250 see the Reactor l Building has pressurized to 38 psig, and the heat transfer from the 4 primary has become limited to the decay heat source. At 360 sec the j energy removal capacity of the RBCS exceeds the decay heat source, and

'l the pressurization of the building has peaked and begins decreasing. The l

peak Reactor Building pressure is significantly less than the design  ;

pressure of 59 psig. No operator action is assumed. This scenario ,

bounds all credible steam line breaks within the Reactor Building."

In regard to the analyses, the Licensee stated (4):

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TER-C5506-134 "The existing FSAR analysis summarized above did not explicitly address the impact of the runout flow from the auxiliary feedwater system. This is because continued feedwater addition to the affected steam generator by means of the main feedwater system is considered to be more limiting with respect to containment pressure response than the case involving '

auxiliary feedwater flow.

The auxiliary feedwater system currently in use at Oconee consists of a turbine driven pump and two motor driven pumps. Each steam generator can receive auxiliary feedwater flow from one motor-driven pump and the

-; turbine driven pump. Operation of the auxiliary feedwater system at ruriout conditions would result in approximately 2050 gym (700 gym from the motor dri.ven pump and 1350 gym from the turbine driven pump) auxil-l iary feedwater addition into the affected steam generator. Since the l

flow capacity of the auxiliary feedwater system is less than that of the sain feedwater pump (greater than 10,000 gpa), since the auxiliary feedwater temperature (90*F) is less than that of the main feedwater (4 60*F) and since the existing analysis considered the maximum possible l cooldown of the primary system for a steam line break in the containment, it is conicuded that the existing analysis of containment pressure l response bounds the situation involving flow from the auxiliary feedwater l

system." I.

l In regard to the ability of the main feedwater (MFW) and emergency ,

feedwater (EFW) pumps to remain operable during a MSIB, the Licensee stated (4]:  ;

" Excessive feedwater delivery (runout flow) from the MFWS is prevented by l the high steam generator level trip of the MFW pumps and would be limited l by the self-limiting heat transfer processes. Excessive feedwater delivery from the EFWS is prevented by the level control system and would have a minimal impact on the building pressure response due to the low enthalpy of the fluid and the low flow capacity of the EFWS in comparison to the MFWS."

3.1.2 Evaluation The Licensee's submittals (3, 4] concerning the containment pressure

. response following a MSLB and applicable sections of the Oconee Nuclear Station FSAR (5) were reviewed in order to evaluate whether the following portions of the acceptance criteria were mets o Criterion 1.a - Continuation of flow to the affected steam generator o Criterion 1.b - Potential for containment overpressure nklin Resea

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f TER-C5506-134 o Criterion 1.c - Ability to detect and isolate the damaged steam generator o Criterion 4 - Potential for AFW pump damage o criterion 5 - Design of steam and feedwater isolation system o criterion 6 - Decay heat removal capacity o Criterion 7 - Safety-grade requirements for MFW and APW isolation valves.

Oconee Nuclear Station Units 1, 2, and 3 are virtually identical, Babcock and Wilcor-designed, two-loop, 860 MWe plants.

In the event of a MSIB, the following systems actuate to provide necessary protections o Reactor trip on high flux (104.9%, two out of four channels) or on low reactor coolant system (RCS) pressure (1800 psig, two out of four channels) o The reactor trip signals:

a. turbine stop valves to trip

b. Integrated control system (ICS) to control steam generator level at the minimum level. (Control-grade)

c. ICS to close MFW control valves and startup control valves to each steam generator (control-grade) o High pressure injection (HPI) system is actuated upons

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a. two out of three (2/3) low reactor coolant system pressure signals j (1500 psig)

b. 2/3 high reactor building pressure signals (4 psig) i o Iow pressure injection (LPI) system is actuated upon

a. 2/3 low reactor coolant system pressure (500 psig)

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b. 2/3 high reactor building pressure j o Reactor building spray system (RBSS) is actuated on 2/3 very high l reactor building pressure signals (10 psig) i Ib J

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TER-C5506-13 4 o Three reactor building cooling units (RBCU) are actuated at 2/3 high reactor building pressure signals.

Each AM system consists of two motor-driven pumps (450 gpa) and one turbine-driven pump (880 gpm). The motor-driven pumps are normally aligned to supply a single OTSG and the turbine-driven pump is aligned so supply both OTSGs. The flow from one motor-driven pump to the unaffected OTSG is sufficient to ensure that the system heat removal exceeds the minimum level required for decay heat removal af ter a MSLB.

The EN systems of the three units may be cross-connected such that any unit may supply EN to another unit.

The safety-grade steam generator level control system (SGLCS) provides automatic OTSG water level control while the EN system is supplying feedwater to the steam generators. SGLCS is designed to automatically control and modulate EN supply to the steam generators during all initiating conditions for the EN system. Each OrSG has two independent level control systems each of which is capable of supplying a signal to the OTSG EN level control valve. All automatic initiation logic and control functions are independent of the ICS. .

The environmental qualification of safety-related electrical and mechanical components is being reviewed separately by the NRC and is not within the scope of this review.

The above systems are designed to safety-grade and IEEE Std 279-1968 requirements. The compliance of these systems with IEEE Std 279-1971 requirements was not reviewed.

The review did not determine whether the instrumentation that the operator relies upon to follow the accident and isolate the affected steam generator conforms with the criteria in ANS/ ANSI 4.5-1980 and Regulatory Guide 1.97.

The worst-case MSLB is a double-ended rupture, at full rated power, with no operator action to isolate MN. Water level in the affected 0 $G is -

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assumed to be maintained at the 2-f t minimum level by MN. At 250 sec, the reactor building pressure reaches 38 psig; this amounts to a back pressure on

< the system which limits RCS temperature to a minimum of 248*, thus halting the i

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a TER-C5506 -13 4 sensible heat flow from the coolant system. Beyond this time, only decay heat is released. The mass and energy released from this time until the time when the energy removal rate exceeds the decay heat generation rate is less than that required to reach building design pressure (59 psig) .

The heat removal rate of the RBCUs equals the decay heat rat'e at 6 min following the MSLB. Each of two RBSS trains provides an energy removal capacity of 120 x 10 Btu /hr and each of three RBCUs provides an energy removal capacity of 80 x 10 Btu /hr. The arrangement of the engineered safeguards power supplies ensure that at least one RBSS train and two RBCUs are available in the event of a fault on a bus, thus providing a minimum 6

energy removal capacity of 280 x 10 Btu /hr.

The EN pumps are protected from damage caused by operating at runout flow by the SGLCS which throttles the EN flow to the OTSGs.

3.1.3 Conclusion and Recommendations The Licensee's responses (3, 4] and the Oconee Nuclear Station FSAR [5]

adequately address the concerns of Item 1 of IE Bulletin 80-04. The containment pressure response analysis and the design of the engineered safeguards satisfy the NRC's acceptance criteria. Regarding Item 1, it is concluded that there is no potential for containment overpressurization resulting from a MSLB with continued feedwater addition. The E N pumps are adequately protected against a i

runout flow condition and therefore will be able to carry out their intended function without incurring damage in the event of a MSIA.

4 3.2 REVIN OF REACTIVITY INCREASE ANALYSIS The requirement from IE Bulletin 80-04, Item 2, is as follows:

" Review your analysis of the reactivity increase which results from a main steam line break inside or outside containment. This review should consider the reactor cooldown rate and the potential for the reactor to return-to-power with the most reactive control rod in the fully withdrawn position. If your previous analysis did not consider all potential water sources (such as those listed in 1 above) and if the reactivity increase is greater than previous analysis indicated, the report of this review should include:

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a. S e boundary conditions for the analysis, e.g., the end of life shutdown margin, the moderator temperature coefficient, power level and the not effect of the associated steam generator water inventory on the reactor system cooling, etc.,

b. The most restrictive single active failure in the safety injection system and the effect of that failure on delaying the delivery of high concentration boric acid solution to the reactor coolant system,

c. The effect of extended water supply to the affected steam generator on the core criticality and return-to-power, I

d. Se hot channel factors corresponding to the most reactive rod in the fully withdrawn position at the end of life, and the Minimum Depar-l ture from Nucleate Boiling Ratio (MDISR) values for the analyzed l

transient."

3.2.1 Susmary of Licensee Statements and Conclusions In regard to the reactivity increase resulting from a MSLB with continued feedwater addition, the Licensee stated (3):

"The Oconee FSAR analysis of the reactivity increase resulting from a .

main steam line break considered four cases involving various potential modes of feedwater addition to the affected steam generator. . In all I cases, a minimum rod worth, based on the maximum worth rod considered

! stuck-out, and consistent with the minimum shutdown margin required by the Technical Specifications was utilized in the core reactivity {'

calculation. The pertinent assumption and system responses for each of these cases are summarized below.

1 Case 1 In this case, the integrated control system (ICS) was assumed to initially close the main and startup feedwater control valves following the reactor trip and then the operator was assumed to maintain feedwater isolation of the affected steam generator. The minimum (two foot) level was maintained in the unaffected steam generator. Under these assump-tions, the reactor was calculated to rensin subcritical throughout the transient.

Case 2 In this case, also the ICS was assumed to close the main and startup feedwater control valves following the reactor trip; however, no credit was taken for operator action to maintain feedwater isolation of the affected steam generator. Consequently, feedwater flow by means of the main feedwater pump continued to the affected steam generator at a rate

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3 TER-C5506-13 4 necessary to maintain a two foot level. The resulting cooldown of the primary system was calculated to cause a return to power of about 1% FP (full powerl at approximately 170 seconds. The core then returned to subcritical conditions with the addition of highly borated water by the emergency core cooling system (HPI, CFT [ core flood tanks], and LPI) .

Case 3 The third steam line break analysis case assumed proper ICS action to initially close the main and startup feedwater control valves; however, no operator action to maintain feedwater isolation of the affected steam generator was assumed. The auxiliary feedwater pump was assumed to start on a low main feedwater pump discharge pressure signal. The ICS was assumed to maintain a minimum (two foot) level in both steam generators with a combination of main and auxiliary feedwater. The analysis predicted a return to 35 percent of rated power in approximately 65 seconds. Without the stuck rod and considering the nominal trip rod worth, the core was found to remain subcritical.

Case 4 The fourth main steam line break analysis case included the assumption of no ICS or operator action to change the feedwater control valve positions. The feedwater flow to the damaged steam generator was postulated to be 135% of the rated flow in one steam generator. It was assumed that the auxiliary feedwater system was not actuated. Under

, these conditions, the reactor was calculated to return to less than 8 percent of rated power approximately 166 seconds after the break before going suberitical again by injection of borated water by the ECCS

[ emergency core cooling system] .

From the foregoing discussion, it is seen that the existing analysis of the steam line break accident considered several potential modes of p feedwater addition to the affected steam generator from the main and 4 auxiliary feedwater systems. Although the flow capacity of the auxiliary feedwater system has increased with the recent addition of the motor

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j driven pump into each of the two secondary loops, the analysis for Case 3  ;

l above still represents the worst case core reactivity increase. Since O the increases in the auxiliary feedwater flow capacity is very small l l (less than 54) compared to the available total feedwater flow capacity j l and since the amount of feedwater flow into the steam generator is i dictated by the steam generator level requirement."

3.2.2 Evaluation

! The Licensee's analysis of the core reactivity increase resulting from a i

MSLB with continued feedwater addition was reviewed in order to evaluate whether the following acceptance criteria were mets >

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TER-C5506-13 4 o Criterion 1.c - Ability to detect and isolate the damaged steam generator o Criterion 1.d - Changes in core reactivity increase o criterion 3 - Analysis assumptions.

The FSAR analysis of the reactivity increase resulting from a MSIB and Reference 3 were reviewed. From that review, it was determined that the analysis is conservative in its assumptions and that the assumptions are in ~-

accordance with those in Acceptance Criterion 3.

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In the worst case MSLB, which assumes full power conditions, a double-ended rupture at the steam generator exit and no operator action to isolate MFW and the ICS is assumed to actuate the turbine-driven EPW pump. The reactor returns to a peak power of 35% at 65 sec and then returns to suberiticality. The calculated return-to-power does not result in a violation of the specified acceptable fuel design limits.

3.2.3 conclusion .

The Licensee's responses (3, 4] and FSAR [5] adequately address the concerns of Item 2 of IE Bulletin 80-04. All potential sources of water were identified in the FSAR analysis, and although a reactor return-to-power is predicted, the specified acceptable fuel design limits are not exceeded. The FSAR analysis of the reactivity increase resulting from a MSLB remains valid. j 3.3 REVIEW OF CDRRECTIVE ACTIONS The requirement from IE Bulletin 80-04, Item 3 is as follows:

"If the potential for containment overpressure exists or the reactor return-to-power response worsens, provide a proposed corrective action and a schedule for completion of the corrective action. If the unit is l-operating, provide a description of any interim action that will be taken until the proposed corrective action is completed."

3.3.1 Susumary of Licensee Statements and Conclusions The Licensee stated (3):

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i TER-C5506-13 4 "As demonstrated in the response to Item 1 above, the potential for containment overpressure is not introduced by postulated auxiliary feedwater pump operation at runout conditions. Furthermore, the existing emergency procedure includes operator guidance to prevent uncontrolled

feedwater addition to the affected steam generator. S e reactor return-to-power responses calculated in the FSAR still represent the limiting case for core reactivity increase. Therefore, no corrective actions are considered necessary at this time for oconee Nuclear Station. It is pointed out that a probabilistic risk assessment study is being planned for Oconee. If the results of this study, indicate the need for any corrective actions with respect to the steam line break accident, appropriate corrective actions will be considered at that time."

3.3.2 Evaluation and Conclusion The Licensee's analysis determined that a containment overpressurization or a worsening of a reactor return-to-power with a resultant violation of specified acceptable fuel limits resulting from a MSIA would not occur.

Therefore, it is concluded that no further action regarding IE Bulletin 80-04 is required of DPC for the Oconee Nuclear Station Units 1, 2, and 3.

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4. CONCLUSIONS With respect to the Oconee Nuclear Station Units 1, 2, and 3, conclusions regarding Duke Power Company's response to IE Bulletin 80-04 are as follows:

o There is no potential for containment overpressurization resulting from a main steam line break (Msta) with continued feedwater addition. .;*

o The emergency feedwater (EN) pumps are adequately protected against a runout flow condition and therefore will be able to carry out their intended function without incurring damage in the event of a MSLB.

o All potential water sources were identified and, although a reactor j return-to-power is predicted, the specified acceptable fuel design limits are not exceeded; therefore, the FSAR reactivity increase analysis remains valid.

o No further action is required by the Licensee regarding IE Bulletin r 80-04.

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5. REFERENCES

. 1. " Analysis of a PWR Main Steam Line Break with Continued Feedwater Mdition" NHC Office of Inspection and Enforcement, February 8,1980 IE Bulletin 80-04

2. "Overpressurization of the Containment of a PWR Plant af ter a Main Line Steam Break" NBC Office of Inspection and Enforcement, October 1,1979 IE Information Notice 79-24 3.

W. O. Parker, Jr. (DPC)

Inttee to J. P. O'Reilly' (NRC, Region II)

Subject:

IE Bulletin 80-04 May 7, 1980

4. W. O. Parker, Jr. (DPC)

Intter to H. R. Denton (NRR)

Subject:

Response to Request for Mditional Information, PWR Main Steam Line Break with Continued Ptedwater Mdition Oconee Nuclear Station Units 1, 2, and 3 l

July 23, 1982

5. Oconee Nuclear Sta' tion Final Safety Analysis Report Duke Power Company, 1982

6. Technical Evaluation Report "PWR Main Steam Line Break with Continued Feedwater Mdition - Review of Acceptance Criteria" Franklin Research Center, November 17, 1981 TER-C5506-119

7. " Criteria for Protection Systems for Nuclear Power Generating Stations" Institute of Electrical and Electronics Engineers, New York, NY,1971 IEEE Std 279-1971

8. Standard Review Plan, Section 4.2

" Fuel System Design" NRC, July 1981 NUREG-0800 1

9. Standard Review Plan, Section 15.1.5 l

" Steam System Piping Failures Inside and Outside of Containment (PWR) "

NRC, July 1981 NUREG-0800 ,

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10. " Criteria for Accident Nonitoring Punctions in Light-mter-cooled Reactors

• American Nuclear Society, Hinsdale, IL, December 1980 ANS/ ANSI-4.5-1980 i

11. " Instrumentation for Light-Water-Cooled Nucle &r Power Plants to Assess Plant and Environs Conditions During and FOllowing an' Accident" Rev. 2 NBC, December 1980

• Regulatory Guide 1.97

12. " Single Failure Criteria for PWR Fluid Systems" American Nuclear Scciety, Hinsdale, IL, June 1976 ANS-51.7/N658-1976

13. " Quality Group Classifications and Standards for Water, Steam, and Radioactive-Weste-Containing Q)mponents of Nuclear Power Plants" _

Rev. 3 NBC, February 1976 lI

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Regulatory Guid. 1.26

14. "Interia Staff Position on Environmental Qualification of e p

i Safety-Related Electrical Equipment" l Rev. 1 NRC, July 1981 NUREG-0588 '

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