ML20069C592

PWR Main Steam Line Break W/Continued Feedwater Addition (B-69),Toledo Edison Co,Davis-Besse Nuclear Power Station Unit 1, Technical Evaluation Rept

ML20069C592
Person / Time
Site:	Davis Besse
Issue date:	06/23/1982
From:	Vosbury F FRANKLIN INSTITUTE
To:	Peter Hearn NRC
Shared Package
ML20069C595	List: ML20069C592
References
CON-NRC-03-81-130, CON-NRC-3-81-130 IEB-80-4, TAC-46832, TER-C5506-128, NUDOCS 8206280298
Download: ML20069C592 (20)

Text

N$$k 1

TECHNICAL EVALUATION REPORT PWR MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION (B-69)

TOLEDO EDISON COMPANY DAVIS-BESSE NUCLEAR POWER STATION UNIT 1 xx NRC DCCKET NO. 50-346 FRC PROJECT C55C6 NRC TAC NO. 46832 FRC ASSIGNMENT 5 NRC CONTRACT NO. NRC43 81 130 FRC TASK 12S Prepared by y, q, yn,yun Franklin Research Center Author: S. M. Jenkins The Parkway at Twentieth Street M. A. Fedele Philadelphia, PA 19103 FRC Group Leader: R. C. Herrick Prepared for Nuclear Regulatory Commission Washingtnn, D.C. 20555 Lead NBC Engineer- P. Hearn June 23, 1982 This report was prepared as an account of work sponsored by an agency of the United States Govemment. Neither the United States Govemment nor any agency thtreof. cr any of their employees, makes any warranty, ex-pressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product or process disclosed in this report, or recretents that its use by such third party would not infringe privately owned rights.

6 "W f

.t . Franklin Research Center ADivision oTThe Franklin Insutute m a- re am.m it inics aiwa icco i OmnC290'820623 CF ADOCK OCOOO346 .

$d5 , n .

1 1

)

l TER-C5506-128 CONTENTS Section Title Page 1 INTRODUCTION. . . . . . . . . . . . . . 1 1.1 Purpose of Review . . . . . . . . . . . 1 1.2 Generic Background . . . . . . . . . . . 1 1.3 Plant-Specific Background . . . . . . . . . 3 2 ACCEPTANCE CRITERIA . . . . . . . . . . . . 4 3 TECHNICAL EVALUATION. . . . . . . . . . . . 8 3.1 Review of Containment Pressure Response Analysis . . . 8 3.2 Review of Reactivity Increase Analysis . . . . . . 12 3.3 Review of Corrective Actions . . . . . . . . 13 4 CONCLUSIONS . . . . . . . . . . . . . . 15 5 REFERENCES . . . . . . . . . . . . . . 16 l

l l

1 l

[

iii I

&[ nklin Jd

~~ Resea_rch Center

I TF.R-C5506-128 FOREWohD Tais Technical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. Nuclear Regulatory Com:sission (Office of Nuclear Reactor Regulation, Division of Operating Reactors) for technical assistance in support of NRC operating reactor licensing actions. The technical evaluation was conductea in accordance with criteria established by the NRC.

Mr. F. 4. Voscury and Mr. S. M. Jenkins contributed to the technical preparation of this report through a subcontract with WESTEC Services, Inc.

Mr. M. A. Fedele contributed to the preparation of this report through a succontract with Evaluation Associates, Inc.

JUh nklin Research Center m m v~ ~

I TER-C5506-128

1. INTRODUCTION 1.1 PURPOSE OF REVIEW Tnis Technical Evaluation Report (TER) documents an independent review of the Toledo Edison Company's response to the Nuclear Regulatory Commission's (NRC) IE Bulletin 80-04, " Analysis of a Pressurized Water Reactor Main Steam Line Break witn Continued Feedwater Addition" (1), as it pertains to the Davis-Besse Nuclear Power Station Unit 1. This evaluation was performed with the following objectives:

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

1.2 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, ano it was cetermined 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 AFW system had not been consicered in the earlier analysis.

On Octooer 1, 1973, the foregoing information was provided to all holders of operating licenses and construction permits as IE Information Notice 79-24

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

4 200U Franklin Resee.rch Center A Omaean af The Fremman meanne

i l

TER-C5506-128 A third licensee informed the NRC of an error in the MSIA analysis for

(

tneir plant. During a review of the MiIA adsalysis, for zero or low power at the end of core life, the licensee identified an incorrect postulation that the startup feenwater control valves would remain positioned "as is" during 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.

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

"1. 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 feeowater 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 i llity of the pumps to remain operable after extended operation at runout flow.

2. Review your analysis of the reactivity increase which results from a main steam line break inside or outside containmenc. 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 tully 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 tne report of this review should include

a. The bouncary conditions for the analysis, e.g., the end of life shutoown 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 reacter coolant system, MMJ7ranidinResearch Center A On=en cf The Fewmen sumaan

EM:E2Bi+FsEf*

1 i TER-C5506-128

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

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 Nucleate Boiling Ratio (MDNBR) values for the 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 completion 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."

1.3 PiaudT-SPECIFIC BACKGROUND Toleco Eoison respondad to IE Bulletin 80-04 in a letter to the NRC dated May 5, 1980 (3). The information in Reference 3 and pertinent information trom the Davis-desse Final Safety Analysis Report (FSAR) [4] were evaluated to cetermine the adequacy of the Licensee's response to IE Bulletin 80-04.

i 2.5 Frankun Research Center A Opamen af The Pm -

TER-C5506-128

2. ACCEPTANCE CRITERIA The following criteria against wnich the Licensee's MSLB response was evaluated were provided by the NaC [5]:

1. PWR licensees' responses to IX Bulletin 30-04 shall include the following information related to their analysis of containment pressure ano core reactivity response to a MSLB within or outside containment:

a. A discussica of the continuation of flow to the affected steam generator, including the impact of runout flow from the AM system and the impact of other energy sources, such as continuation of feedwater or concensate flow. AN system runout flow should be determined from the manufacturer's pump curves at no backpressure, unless the system contsins 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 components (e.g.,

i automatically throttled valves) which meet the requirements of l IEEE Std 279-1971 [6] and passive devices (e.g., flow orifices or cavitating venturis).

b. A determination of potential containment overpressure as a result of the impact of runout flow from the AFW 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 maae 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 Gd the ability to detect and isolate the damaged steam generator from continued feedwater addition during the MSLB  ;

acciaent. Operator action to isolate APW flow to the affected steam generator within tne first 30 minutes of the start of the ,

MSIR should be justified. The justification should address the indication available to the operator and the actions required, particularly those cutside the control room. If operator action is required to prevent em:eeding a design value, i.e.,

containment design pressure or departure from nucleate boiling ratio (DnBR), then the discussion should include the calculated  !

time when the design value would be exceeded if no operator action were assumed. j i

O ra

!!dh.o nklin

. n. eResear.ch Center

TER-C5506-128

d. Where all water sources were not considered in the previous analysis, an indication should be provided of the core reactivity change which results from tne inclusion of additional water sources. A submittal which does not determina the magnitude of reactivity change from an original analysis is not responsive to the requirements of IE Bulletin 80-04.

2. If the licensee's analysis shows that containment overpressure or a reactor-return-to-power with a DNBR less than 1.32 (1.30 for Tong correlation) (*) can occur, then the licensee shall provide the following adcitional informations

a. The proposed corrective actions to preclude overpressure or reactor-return-to-power and a schedule for completion of those actions.

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

3. Tne 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 (7) . The 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 conoucted a sensitivity analysis which demonstrates that a particular assumption is more conservative.

Assumption II.3.d.: The most restrictive single active failure in the safety injection systen 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.: The initial core flow snould be chosen such that the post-MSLB shutdown margin is minimized (i.e., maximum initial core flow) .

• 0tner values for minimum DNBR may be acceptaDie if justified for certain fuel designs and DNBR correlations.

dbranklin Research Center A Daumen ad The Famuses ammmane

i .

J TER-C5506-128 Se acceptaole computer codes for the licensee's analysis of core i reactivity changes are, by nuclear steam supply system (NSSS) vendor, the followings CESEC (CE) , IDFTRAM (Westinghouse) , and TRAP (B&W) .

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 ocaputer 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 AN pumps can be canaged 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 snoula be indicated to the NHC for furcher action and resolution.

5. heifications to the electrical instrumentation and controls needed to detect and initiate isolation of the effected ateam generator ano feedwater sources in order to prevent containment overpressure and/or unacceptable core reactivity increases must satisfy safety-grade requirements. Instrumentation that the operator relies upon to follow the accident and to determine isolation of tne affected steam generator and feeowater sources should conform to the criteria contained in ANS/AMSI-4.5-1980, " Criteria for Accident Monitoring Functions in I.ight-Mater-Cooled Reactors" [8), and the regulatory positions in Regulatory Guide 1.97, Rev. 2, "Instrumentatien for Iaght-44ater-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and M11owing an Accident" [9].

6. Auxiliary feedwater system status should be reviewed to ensure that system heat rear /;al 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 wnich adversely affect vital assumptions of the containment pressure and core reactivity response analyses.

7. Se safety-grade requirements (redundancy, seismic and environmental qualifications, etc.) of the equipment that isolates the main feedwater (MN) and AN systems from the affected steam generator snoula be specified. S e modifications of equipment relied upon to isolate the MN and AN systems from the affected steam generator should satisfy the following criteria to be considered safety-grade:

o nee ndancy and power source requirements: 2e isolation valves i snoula be designed to acea==ah te a single failure. A failure- l noces-and-effects analysis should demonstrate that the system is capable of withstancing a single failure without loss of function. i he single failure analysis should be conducted in accordance with

[ranWin

_ _Research Center

TER-C5506-128 the appropriate rules of application of ANS-51.7/N658-1976,

" Single Failure Criteria for PWR Fluid Systems" (10) .

o Seismic requirements: The isolation valves should be designed to Category I as recommended in Regulatory Guide 1.26 [11).

o Environmental qualification: The isolation valves should satisfy the requirements of NUREG-05d8, Rev.1, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment" (12).

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.

i I

l l

l dbu Franklin Research Center A > w w n an m

TER-C5506-128

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

1. Review the 7.icensee'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 contir ued 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 j information presented for Tasks 1 and 2 above.

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

3.1 REVIEW OF CONTAINNENT PRESSURE RESPCNSE 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 af 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 your acility to ostect aan isolate the damaged steam generator from these sources and the ability of the pumps to remain operable af ter extended operation at runout flow." ,

a. Summary of Licensee Statements and Conclusions The Licensee made the following statements regarding t!.e ability to detect and isolate a MSIA:

~

[ranklin Research Center

TER-C5506-128 l "2o11owing a postulated double ended rupture of a main steam line inside the containment vessel, low pressure switches locatad in the main steam line outside containment will actuate the Steam and Feedwater Rupture Control System (SFRCS). The reactor trips on low reactor coolant system pressure. SFRCS will trip the turbine-generator, initiate cicsure of the  ;

main steam isolation valves in both main steam lines, initiate main 1 feedwater isolation for both steam generators, and will start the auxiliary feedwater system.

"SFRCS will determine which is the affected steam generator. No auxiliary fewdwater will be added to the affected steam generator, as SFRCS will align both auxiliary feedwater pumps to supply water only to the unaffected steam generator.

"Following main steam and main feedwater isolation of the steam generator, the unaffected steam generator will repressurize while the affected steam generator will continue to blow down and will not repressurize. The redundant SFRCS pressure switches on the main steam lines of each steam generator (set of 600 psig) are the means of detecting which is the affected generator. These pressure switches are interlocked with the valves in the auxiliary feedwater system to prevent the addition of auxiliary feedwater to the affected steam generator."

In regard to the. review of the containment pressure response analysis for the Davis-Besse plant, the Licensee stated:

"During that period of time required to detect MSLB and close the main steam and feedwater valves, main feedwater will continue to flow to both the affected and the unaffected steam generators. This feedwater addition has been considered in the MSIA analysis and is described in the response to FSAR Questioa 15.4.8 and is tabulated in FSAR Table 15.4.4.2. We have reviewed the data used in the analysis and have found that the values used for feedwater addition following the MSLB are conservative.

"The values used in the analysis exceed the expected feedwater addition that would result from the most adverse response of the non-safety grade portion of the main feedwater system (control valves wide open and main feed pump turbine overspeed) and the most limiting single failure in the safety grade portions of the feedwater isolation system (failure of the main feedwater stop valves to close on the affected steam generator)."

The Licensee concluded: .

"Since our review has determined that the existing FSAR analysis for containment overpressure following MSLB conservatively models continued f eedwater addition, no further action is required."

Regarding the AFW pump's ability to remain operable after extended operation at runout flow, the Licensee stated:

_9 I.u Franklin

% m the r Research un Center

1 TER-C5506-128 "Because the auxiliary feedwater system only supplies water to the unarrected steam generator, there is no runout of the auxiliary feedwater pumps and their continued availability is not affected."

o. Evaluation Tne Licensee's submittal concerning containment pressure response analysis and applicable sections of the Davis-Besse FSAR were reviewed in orcer to ovaluate whether the following portions of the acceptance criteria were mets o Criteria 1.a - Continuation of flow to the affecteo steam generator o Criteria 1.b - Potential for containment overpressure o Criteria 1.c - Ability to detect and isolate the damageo steam generator o Criteria 4 - Potential for AFW pump damage o Criteria 5 - Design of steam and feedwater isolation system o Criteria 6 - Decay heat removal capacity o Criteria 7 - Safety-grade requirements for MFif and AFW isolation valves.

A review of Section 7 of the Davis-Besse FSAR determined that the SFRCS system is designed as an engineered safety features (ESP) system to Seismic Category I ano safety-grace requirements, and the initiating signals and circuits were designed to meet the criteria of IEEE Std 279-1971.

In the event of a MSLB, the SFRCS is designed to o isolate the main steam ana main feedwater system o start both AFW system pumps o align AFW flow to feed only the unaffected steam generator.

Steam generator level then is automatically maintained by the AFW system. The opacator also has the option to take manual control of steam generator level or transfer control to the integrated control system (ICS),

wnica will maintain a wider band of steam generator level.

4Jbj Franknn Research Cemer 4 cm .e the r m

• i 5 i

TER-C5506-128 The environmental qualification of safety-related electrical anc mechanical components is being reviewed separately by the NRC and is not within tne scope of tnis review. The qualfication of the instrumentation that the operator relies upon to follow accident and determine isolation of the affected steam generator was not cetermined.

Sufficient AW flow is available to the unaffected steam generator to enqure that system heat removal capacity exceeds the minimum level required for cecay heat removal after a MSLB.

Review of the steam line break analysis in the FSAR determined that the effects of continued feedwater addition had been adequately addressed. The SFRCS and AW system are designed so that, even if a single failure to either system would occur, AN flow would not reach the affected swam generator.

The worst-case single failure is that ot the MN control valves (control-grade) remaining 100% open, allowing MFW pump runout to 135% full MN flow to the affected steam generator fcr 17 seconds until the feedwater isolation valves (safety-grade) close. The PSAR analysis ceter:sined that the resultant release of aass ano energy from the blowdown of one steam generator would increase the containment pressure to 21.4 psig, well below the design pressure of 36 psig.

The centinued availability of the AN pumps would not be affected since the pumps do not experience runout flow.

c. Conclusions and Recommendations Tne Lacensee's response and FSAR adequately address the concerns of Item 1 of IE Bulletin 80-04. The containment pressure response analysis and the cesign of the SFRCS 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. In addition, since the AN pumps do not experience runout conditions, the pumps will be able to carry out their intended function without incurring damage.

,g dhbrank5n Resee ch Center 4on.on.e m r w m

i l

l l

i i

TER-C5506-128 I

3.2 REVIDi 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 tne 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 snould includes

a. The boundary conditions for the analysis, e.g., the end OZ 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. Tne most restrictive single active failure it 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,

d. The bot channel factors corresponding to the most reactive rod in the fully witndrawn position at the end of life, ano the Minimum Departure from Nucleate Boiling Ratio (MDNBR) values for the analyzed transient."

4. Summary of Licensee Statements and Conclusions In regard to the reactivity increase resulting from a MSLB with continued  ;

feedwater addition, the Licensee stated:

"...the rupture of a main steam line between the steam generator and the main steam isolation valve (MSIV) represents the worst condition for accident 2nMysis. Our review of feedwater addition described in the res@nse to Item 1 of this Bulletin has also considered a break outside tne containment and upetream of the MSIV. A break in this location will result in slightly less feedwater addition due to the proximity of the SFRCS pressure switch taps to the break and the consequent earlier SFRCS trip signal.

"Since our review has determined that existing FSAR analysis of

! reactivity increase following MSLB inside or outside containment considers all potential water sources, and conservatively models those sources, no further action is required."

M JS Franklin Research Center aon .#Ther - a -

MTWM" l

TER-C5506-128

b. Evaluation Tne Licensee's analysis of the core reactivity increase resulting from a MSLB with continued feedwater adoition was reviewed in order to evaluate wnetner tne following acceptance criteria were mets o Criteria 1.c - Ability to detect and isolate the damaged steam generator o Criteria 1.d - Changes in core reactivity increase o criteria 3 - Analysis assumptions.

Review of tne PSAR analysis of the reactivity increase resulting from a MSLB cetermined that the analysis is conservative in its assumptions and that the assumptions are in accordance witn those in Acceptance criteria 3.

As discussed in Section 3.1.b of this report, the SFRCS isolates all potential water sources from the affected steam generator. In the worst case (double-ended MSLB between the steam generator and main steam isolation valve),

no return to criticality occurs, the minimum suberitical margin during the transient is 0.69% Ak/k, and the minimum DNBR achieved during the transient is 1.42.

All potential water sources were considered in the FSAR analysis of the reactivity increase resulting from a MSLB and no further action is required.

c. Conclusion The Licensee's response and FSAR adequately address the concerns of Item 2 of IE Bulletin 80-04. All potential sources of water were identified, the SFRCS isolates 411 the potential water sources, no return-to-power occurs, and the DNBR remains greater than 1.30. Therefore, the FSAR analysis remains valid and no further action is required.

3.3 REVIIM OF CORRECTIVE ACTIONS Tne requirement from Id Bulletin 80-04, Item 3, is as follows:

"I1 the potential for containment overpressure exists or the reactor-return-to-power response worsens, provide a proposed corrective action 4

bbnklin 4 t=sa w Research.

Th. r Center

l TER-C5506-128 and a senedule for completion of the corrective action. If the unit is operating, provide a description of any interim acciar3,that will be taken until the proposed corrective action is completed."

a. Summary of Licensee Statements and Conclusions Tne Licensee stated:

"Since our review has determined that the existing FSAR analysis for l containment overpressure following MSLB conservatively models continued teeowater addition, no further action is required.

"Since our review has determined that the existing FSAR analysis of reactivity increase following MSLB inside or outside containment considers all potential water sources, and conservatively models those sources, no further action is required."

b. Evaluation The Licensee's conclusions with respect to the corrective actions required ,

as a result of the review were evaluated against Acceptance Criteria 2.

The existing FSAR analysis for containment overpressurization following a MSLB takes into account all potential water sources and no further action is

required by the Licensee. ,

I Since all potential water scurces were identified in the FSAR analysis of tne reactivity increase following a MSLB, no further action is required.

c. g nelusion and Recommendations  ;

The Licensee's analysis determined that containment overpressurization or a worsening of a reactor return-to-power with a DNBR of less than 1.30

! resulting from a MSIA would not occur. Therefore, it is concluded that no further action regarding IE Bulletin 80-04 is required of Toledo Edison for the Davis-Besse Nuclear Power Station Unit 1.

i I

i I-

[

db FrankHn Research Center ,

A Oneuen af The Frummm summae

TER-C5506-128

4. CONCLUSIONS With respect to the Davis-Besse Nuclear Power Station Unit 1, conclusions regarding Toledo Edison's response to IE Bulletin 80-04 are as follows:

o There is no potential for containment overpressurization resulting from a MSLB with continued feedwater addition.

o The AFW pumps will not experience runout conditions; therefore, they will be able to carry out their intended function without incurring damage during a MSLB. .

o All potential water sources were identified and no reactor-return-to-power occurar therefore, the FSAR reactivity increase analysis remains valid.

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

i e

i r

8 O

Nbd Franklin Research Center A Deuuman W The Pewmen m

b. WA$

• l s .

9 ,

TER-C5506-128

5. REFERENCES

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

2. Overpressurization of the Containment of a PWR Plant after a Main Line Steam Break NRC Office of Inspection and Enforcement, Octooer 1, 1979 IE Information Notice 79-24

• • • 3. R. P. Crouse (Toledo Edison)

Letter to J. P. Keppler (NRC, Region I%I)

May 5, 1980

4. Davis-Besse Nuclear Power Station Unit 1 Final Safety Analysis Report, througn Rev. 27 '

Toledo Edison Company, August 1977

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

6. " Criteria for Protection Systems for Nuclear Power Generating Stations" Institute of Electrical and Electronics Engineers, N4 w York, NY, 1971 IEEE Std 279-1971

7. Standard Review Plan, Section 15.1.5

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

NRC, July 1981 NUREG-0800

8. " Criteria for Accicent Monitoring Functicns in Light-Water-Cooled Reactors" American Nuclear Society, Hinsdale, IL, December 1980 ANS/ ANSI-4.5-1980

. 9. " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Tussess Plant and Environs Conditions During and Following an Accident" Rev. 2 NRC, Decemoer 1980 Regulatory Guide 1.97

10. " Single Failure Criteria for PWR Fluid Systems" American Nuclear Society, Hinsdale, IL, June 1976 ANS-51.7/N658-1976 Kanklin Researen Center A Dramon of The Fremen evenne

TER-C5506-128

11. " Quality Group Classifications and Standards for Water, Steam, and Racioactive-Waste-Containing Components of Nuclear Power Plants" Rev. 3 NHC, February 1976 Regulatory Guide 1.26

12. " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment" Rev. 1 NRC, July 1981 NUREG-0588

. - ~ . - -

_ _ _ . _ _