ML20066G317
| ML20066G317 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 11/18/1982 |
| From: | Vosbury F FRANKLIN INSTITUTE |
| To: | Peter Hearn NRC |
| Shared Package | |
| ML14190A691 | List: |
| References | |
| CON-NRC-03-81-130, CON-NRC-3-81-130, TASK-B-69, TASK-OR TAC-46856, TER-C5506-137, NUDOCS 8211220133 | |
| Download: ML20066G317 (24) | |
Text
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TECHNICAL EVALUATION REPORT PWR MAIN STEAM LINE BREAK WITH
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CONTINUED FEEDWATER ADDITION (B-69)
'i CAROLINA POWER & LIGHT COMPANY H. B. ROBINSON 'JNIT 2 NRCDOCKETNO. 50-261 FRC PROJECT C5506 NRCTACNO. 468S6 FRC ASUlGNMENT3 NRC CONTRACT NO. NRC441130 FRCTASK 137 f
Ptsporedby Franidin Research Center Author:
F. W. Vosbury 20th and Rees Street Philadelphie, PA 19103 FRC Group Lander:
R. C. Herrick Fisperedfor Nuclear Regulotsry Cw..,rRn Washingmn, D.C. 205ti5 Lead N8tC E.h P. Hearn i
November 18, 1982
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This report was crepared as an acccur:t of work sponsore<.1 by an a0em.y of the Uni ed States Govemment. Neither the United States Government nor any aCency ttiereof, or an/ of their employees. makes any werranty, empressed or impiled. cr assumes any legal liabliity or responsibility for any tfilte party's use, or the rtcults of such use, of any Informatfort, app >
ratus, product or process disclos&J In t".ls report, or represents that its use by sucn third party would not infnnge privately owned dghts.
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PWR MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION (B-69) ff
'i CAROLINA POWER & LIGHT COMPANY F
H. B. ROBINSON UNIT 2 NRCDOCKETNO. 50-261 FRC PROJECT C51106 NRCTACNO. 46856 FRC ASSIGNMENT 5 NRC CONTRACT NO. NRC-0341 130 FRCTASK 137
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!i Preparedby Franklin Research Center Author:. F. W. Vosbury
.s 20th and Race Street i
Philadelphia, PA 19103 FRC Group' Leader:
R. C. Herrick
't Preparedfor Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer:
P. Hearn 8
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November 18. 1982 TNs report was prepared 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-
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retus, product or process disclosed in this report, or represents that Ita ure by auch third party would not infringe privately owned rights.
Y Prepared by:
Reviewed by:
Approved by:
cb 'w Ad'*2l....,,W fj Princloal AutMori Group Leader Department Dl/ector#
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Date: //-/7-62.,
Date: //- /f-f 2 Date: Il-18-1L2 -
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CONTENTS 1
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1 l.1 Purpose of Review 1
1.2 Generic Background.
1 1.3 Plant-Specific Background 3
2 ACCEPTANCE CRITERIA.
4 3
TECHNICAL EVAI,UATION.
S 4
j 3.1 Raview of Cbntainment Pressure Response Analysis 8
4 3.2 Review of Reactivity Increase Analysis.
13 N
3.3 Review of Corrective Actions 16 4
CONCLUSIONS.
17 5
REFERENCES 18 l
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This hchnical Evaluation Report was prepared by Franklin Research Center Ih) 'd under a contract with the U.S. Itaclear Regulatory Commission (Office of y?j;e> Nuclear Reactor Regulation, Division of Operating Reactors) for technical j,*,.,j g
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technical evaluation was conducted in accordance with criteria established by glgpg) k,.h.T' the NRC.
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Mr. F. W. Vbsbury contributed to the technical preparation of this report xy hgg through a subcontract with WESTEC Services, Inc.
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INTRODUCTION yn y 9 W 2,y p
1.1 PURPOSE OF REVIEW
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21s Technical Evaluation Report (TER) documents an independent review of 4
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Carolina Power and Light Company's (CPfrL) response to the Nuclear Regulatory Commission's (NRC) IE Bulletin 80-04, " Analysis of a Pressurised Water Reactor Q% Main Steam Line Break with Continued Ptedwater Addition" [1], as it pertains k
to the H. B. Robinson Steam Electric Plant Unit 2.
Bis evaluation was
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performed with the following objectives:
S c~rj o to assess the conformance of CP&L's main steam line break (MSLB) 4 analyses with the requirements of IE Bulletin 80-04 a
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o to assess CP&L's proposed interim and long-range corrective action
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k plans and schedules, if needed, as a result of the MSIa analyses.
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l.2 GENERIC BACKGROUND y.
In the summer of 1979, a pressurized water reactor (PWR) licensee
$ hh submitted a report to the NRC that identified a deficiency in the plant's F
9) p 3 ;q original analysis of the containment pressurization resulting from a MSIa. A
,h
. reanalysis of the containment pressure response following a MSIB 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 j experienced the steam lie break, containment design pressure would be exceeded
,p in approximately 10 minutes. Tne long-term blowdown of the water supplied by l
~d the A2W system had not been considered in the earlier analysis.
4 On October 1,1979, the foregoing information was provided to all holders of operating licenses and construction permits as IE Information Notice 79-24
[2]. Another facility performed an accident analysis review pursuant to receipt of the information in the notice and discovered that, with offsite l[
electrical power available, the condensate pumps would feed the affected steam
.j generator at an excessive rate. His excessive feed v:.s not previously
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. Ad considered in the plant's analysis of a MSLB accioent.
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A third licensee informed the NRC of an error in the MSta analysis for M-
{t their plant. During a review of the MSLB analysis, for sero or low power at i
i the end of core life, the licensee identified an incorrect postulation that the startup feedwater control valves would remain positioned "as is" during the transient. In reality, the startup feedwater control valves will ramp to 808 full open due to an override signal resulting from the low steam generator pressure reactor trip signal. Reanalysis 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-t e r 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 j.
required all PNRs with operating licenses and certain near-term PNR operating license applicants to perform the following:
"1.
Review the containment pressure response analysis to determine if the i y potential for containment overpressure for a main steam line break i
inside containment included the impact of runout flow from the 3g auxiliary feedwater system and the impact of other energy sources,
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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 cperable af ter extended operation at runout flow.
3, 2.
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 L
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
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the reactivity increase is greater than previous analysis indicated the report of this review should includes d
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1he boundary conditions for the analysis, e.g., the end of life i
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 y
injection system and the effect of that failure on delaying the delivery of high concentration boric acid solution to the reactor coolant system, 1
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Se effect of extended water supply to the affected steam s
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generator on the core criticality and return to power,
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Se hot channel factors corresponding to the most reactive: rod in 7
the fully withdrawn position at the end of life, and the Minimum l
Departure from thacleate Boiling Ratio (NDISR) values for the analysed transient.
3.
If the potential for containment overpressure exists or the reactor return-to w er response worsens, provide a proposed corrective h,.
l 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."
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1.3 PLANT-SPECIFIC BACKGROUND l
CP&L responded to IE Bulletin 80-04 in a letter to the NBC dated May 9,
'}N; 1980 [3] and provided additional information in a letter dated September 3, 1982 [4]. The information in References 3 and 4 has been' evaluated along with
'a pertinent information from the H. B. Robinson Unit 2 Updated Final Safety e
Analysis Report (FSAR) [5] to determine the adequacy of the Licensee's
.g compliance with IE Bulletin 80-04.
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ACCEPTANCE CRITERIA The following criteria against which the Licensee's MSIA response was h
evaluated were provided by the NRC [6):
1.
PWR licensees' responses to II Bulletin 80-04 shall include the following information related to their analysis of containment pressure and core reactivity response to a MSIA within or outside containment:
I a.. A discussion 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 condensate flow. A N system runout flow should be
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determined from the manufacturer's pump curves at no backpressure,
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unless the system contains reliable anti-runout provisions or c more representative backpressure has been conservatively calcu-4g,
lated.
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 i
that use active components (e.g., at.tomatically throttled valves) which meet the requirements of IEEE Std 279-1971 [7] and passive devices (e.g., flow orifices or cavitating venturis).
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b.
A determination of potential containment overpressure as a result of the impact of runout flow from the AN 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 exinting FSAR analysis, the analysis must show that runout AN flow was included and that design containeent pressure was not exceeded.
c.
A discussion of the ability to detect and isolate the damaged steam generator from continued feedwater addition during the M812 accident. Operator action to isolate AN flow to the affected steam generator within the first 30 minutes of the start of the MSLB should be justified. If operator action is to be completed I
within the first 10 minutes, then the justification should address I
the indication available to the operator and the actions required.
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Where operator action is required to prevent exceeding a design M
value, i.e., containment design pressure or specified acceptable ^
k fuel design limits, then the discussion should include the calcu-lated time when the design value would be exceeded if no operator action were assumed. Where operator actions ~are to be performed between 10 and 30 minutes af ter the start of the MSLB, the justi-fication should address the indications available to the operator and the operator actions required, noting that for the first 30 3Q.1 minutes, all actions should be performed from the control room.
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Where all water sources were not considered in the previous analysis,
@.'M an indication should be provided of the core reactivity change which plW j results from the inclusion of additional water sources. A submittal 0
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which does not determine the magnitude of reactivity change from an F '
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original ar.alysis is not responsive to the requirements of IE Bulletin 80-04.
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2.
If containment overpressure or a worsening of the reactor return-f.
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to-power with a violation of the specified acceptable fuel design
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1 limits described in Section 4.2 of the Standard Review Plan [8]
E (i.e., increase in core reactivity) can occur by the licensee's 4
analysas, the licensee shall provide the following additional 4
informations q
g a.
the proposed corrective actions to prevent containment.
t overpressure or the violation of fuel design limits, and the
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schedule for their completion b.
the interim actions that will be taken until the proposed 9
corrective action is completed, if the unit,is operating.
4, y
Se acceptable input assumptions used in the licensee's analysis of 3.
the core reactivity changes during a MSTA are given in Section 15.1.5 d
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of the Standard Review Plan (9]. Se following specific assumptions r
should be used unless the analysis shows that a different assumption a
j is more limiting:
Assumption II.3.b. :
Analysis should be performed to determine the
,j most conservative assumption with respect to a
?% g loss of electrical power. A reactivity analysis should be conducted for a normal
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power situation as well as a loss of offsite l
power scenario, unless the licensee has j
previously conducted a sensitivity analysis 4
y which demonstrates that a particular J,
,k assumption is more conservative.
t Assumption II.3.d.:
h e most restrictive single active failure in
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N the safety injection system which has the 1
effect of delaying the delivery of high 3
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concentration boric acid solution to the i
b.g.g reactor coolant syst.em, or any other single l
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active failure affecting the plant response, '
h should be considered.
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pp{. q Assumption II.3.g.:
S e initial core flow should be chosen such h@y'.thf minimized (i.e., maximum initial core flow).
that the post 44STA shutdown margin is
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G'M Se acceptable computer codes for the licensee's analysis of core
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reactivity changes are, by nuclear steam supply system guSSS) vendor, i ' M<,5 the following: CESEC (Combustion Engineering), IDFTRAN (ifesting-Ky;3 house), and TRAF (Babcock & Wilcos). Other computer codes may be y%R used, provided that these codes have previously been reviewed and M;h found to be acceptable by the NRC staff. If a computer code is used i6 which has not been reviewed, the licensee must describe the method
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hj employed to verify the code results in sufficient detail to permit e1 VI the code to be reviewed for acceptability.
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If the AFW pumps can be damaged by extended operation at runou*; flow, j;N l the licensee's action to preclude damage should be reviewed for
'/
technical merit. Any active features should satisfy the requirements M@j%y y of IEEE Std 279-1971. Where no corrective action has been proposed, this should be iridicated to the NRC for further action and resolution.
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Modifications to electrical instrumentation and controls needed to NH detect and initiate isolation of the affected steam generator and Ni feedwater sources in order to prevent containment overpressure and/or MY unacceptable core reactivity increases must satisfy safety-grade ifg requirements.
Instrumentation that the operator" relies upon to F..
follow the accident and to determine isolation of the affected steam
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generator and feedwater sources should conform to the criteria kg contained in ANS/ ANSI-4.5-1980, " Criteria for Accident Monitoring yj Functions in Light-Water-Cooled Reactors" (10), and the regulatory
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positions in Regulatory Guide 1.97, Rev. 2, " Instrumentation for W9 Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs
.h Conditions During and Following an Accident" [11].
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APW syLtem status should be reviewed to ensure that system heat p
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removal capacity does not decrease below the minimum required level
.h ' Q as a result of isolation of the affected steam generator and also
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[f(y affect vital assumptions of the containment pressure and core reactivity response analyses.
7.
Se safety-grade requirements (redundancy, seismic and environmental p ci qualifications, etc.) of the equipment that isolates the main 4
feedwater (MFW) and AFW systems from the affected steam generator
,,([5 should be specified. The audifications of equipment that is relied q
upon to isolate the MFW and AFW systems from the affected steam generator should satisfy the following criteria to be considered
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&yd safety-grades h.T: s d
o Redundancy and power source requirements: The isolation valves 16? ;
should be designed to accommodate a single failure. A failure-Ki modes-and-effects analysis should demonstrate that the system is M{
capable of withstanding a single failure without loss of function.
M9 The single failure analysis should be conducted in accordance h r,?
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" Single Failure Criteria for PWR Fluid Systems" (121 "f,,qg
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o Seismic requirements: The isolation valves should be designed to Category I as recommended in Regulatory Guide 1.26 (131
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Environmental qualifications The isolation valves should satisfy
,[f!$j the requirements of NUREG-0588, Rev.1, "Interia Staff Position
'v(/pdj on Environmental Qualification of Safety-Related Electrical v
Equipment" (141.
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Quality standards: The isolation valves should satisfy Group B iQ quality standards as recommended.in angulatory Guide 1.26 or
{f.[f similar quality standards from the plant's licensing bases.
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TBCENICAL EVALUATION w
'Nh c: y p9 Under contract to the NBC, the scope of work included the following:
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Review the Licensee's response to IE Bulletin 80-04 against the
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acceptance criteria.
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2.
a.
Evaluate the Licensee's MSIA analyses for the potential of j
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overpressurizing the containment and with respect to the core NMi reactivity increase due to the effect of continued feedwater flow.
l b.
Evaluate the Licensee's. Proposed corrective acticas and schedule for implementation if the findings of Task 2a indicate that a V,-
potential exists for overpressurizing the containment or 7
t worsening the reactor return-to-power in the event of a MSLB
-fg'4 accident.
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M 3.
Prepare a TER for each plant based on the evaluation of the f
'y information presented for Tasks 1 and 2 above.
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This report constitutes a TER in satisfaction of Task 3.
Sections 3.1
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' g,
%, c,) through 3.3 of this report state the requirements of IE Bulletin 80-04 by 4
?; subsection, suimmarize the Licensee's statements and conclusions regarding i
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a-f.y
' these requirements, and present a discussion of the Licensee's evaluation t
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3.1 g
REVIEN OF CONTAllGEENT PRESSURE RESPONSE ANALYSIS Ee t
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The requirement from IE Bulletin 80-04, Ito 1, is as follows:
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" Review the containment pressure response analysis to determine if the WI potential for contairument overpressure for a main steam line break inside 3 h:
containment included the impact of runout flow from the auxiliary 7pg.;
feedwater system and the impact of other energy sources, such as t
"'5 continuation of feedwater or condensate flow.
In your review, consider your ability to detect and isolate the damaged steam generator from these y,
sources and the ability of the pumps to remain operable after extended
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operation at runout flow."
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wN 3.1.1 Snamary of Licensee Statements and conclusions
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%p.h In regard to the review of containment pressure response analysis, the Licensee stated [3]:
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>w g,lj h "Se conservative assumptions used in the analysis for containment wm pressure following a main steam line break in containment have been J' N reviewed as requested in IE Bulletin 80-04.
S e analysis documented in
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M 7!il E. B. Robinson Unit 2 [ original] FSAR, Page 14.25-10 included allowance N
for 100 seconds of auxiliary feedwater flow. We have extended the NN.@9 an& lysis to consider auxiliary feedwater flow for 10 minutes, as well as ff main feedwater flow for 10 seconder a conservative estimate of the time 6
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for isolation of the system. Se resultant containment pressure, YW,-
including allowance for auxiliary feedwater flow to 10 minutes, is 34.4 i'l.k}G
$7 peig compared with a design value of 42 psig...."
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MM{1 A main steam line break in containment will result in blowdown to
,ht containment with resultant increase in containment pressure and increase M E'(V in cooling of the RCS. Increased cooling of the BCS would lead to low or b ~ T.
falling pressurizer pressure and level. Se operator would make the M,.h determination of a steam line break on the basis of abnormally low steam
! (fM 3 pressure in one or more steen generators; a continuously decreasirg g
T.,9 would also indicate a steam line break."
In regard to operator action, the Licensee stated [4]:
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"We operator tasks required to identify the affected steam generator and
{'ki?.?. j isolate the AFW flow are taken fros 1&aergency Instruction (EI)-1 Appendix M
B and are as follows:
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Verify that steamline isolation has occurred. If not, manually t;hqq initiate steamline isolatien.
3W Mcl}g 2.
Verify the steam dump valves and stacepheric relief valves are closed
/ j to insure that the emergency has not resulted from an inadvertant
? 3 opening of these valves.
k' d-a 3.
If the reactor coolant pressure drops below 1300 psig, trip all 4 'd reactor coolant pumps after safety injection pump operation is
'AN verified.
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Determine if one steam generator has blowdown by observation of steam
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pressure and isolate the auxiliary feedwater flow to that steam generator.
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Se plant operations staff has evaluated the time required to respond to f*
this accident and has determined that a trained operator responds in 2-3 f.'@4 minutes. Se simulator trainir.g staff has also evaluated the response p.W y]
time for this event and has determined that a typical operator trainee WM response time is 2-5 minutes."
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In response to a request for information regarding AFW flow rate, the I
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Licensee stated [4]:
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"In the original analysis, the back pressure value was conservatively
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calculated for the AN pumps in the runout condition. The back pressure
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value calculation considered elevation differences and line resistance between the AN pump's outlet and the steam generator inlet. The steam gJ generator pressure was assumed to be atmospheric, i.e., sero gage. The 7f"*; 'l runout flow at this pressure for each motor driven AN pump is 316 gym, F'
as calculated in the original analysis."
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iF;;.s:N In regard to the ability of the AN pumps to operate without sustaining damage during a MSIA, the Licensee stated (3]:
Ph Va.C "m difficulties are anticipated with extended auxiliary feed pump FNig operation at runout conditions. Cavitation is not expected at the d'Wf anticipated flow rate."
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3.1.2 Evr.luation y
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i G
f,a The Licensee's submittals (3, 4] concerning the containment pressure 9
g..,9
' j, f.
response following a MSLB and applicable sections of the H. B. Robinson
/ ylaupdated FSAR (5] were reviewed in order to evaluate whether the following
. dR MW.:<l portions of the acceptance criteria were mets
.-Dr.t,..
,,Q 0 gp[
o Criterion 1.a - Continuation of flow to the affected steam generator
] ff
'W't}.4 o Criterion 1.b - Potential for containment overpressure o Criterion 1.c - Ability to detect and isolate the damaged steam
~
P generator o Criterion 4 -
Potential for AN pump damage D{m'
- mg.3 o Criterion 5 -
Design of steam and feedwater isolation sistem E
o Criterion 6 -
Decay heat removal capacity i n o Criterion 7 -
Safety-grade requirements for NN and AN isolation j
,q (g f valves.
h'@[
The H. B. mbinson Unit 2 is a Westinghouse-d asigned, 3-loop, 23004 cit plant.
ii %
(
The following systems provide the necessary protection against a steam
[4 pipe ruptures w
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o Safety injection system actuation ont c
.- 1
.g a.
two out of three low pressuriser pressure signals suf F:M/
b.
two out of three differential pressure signals between any steam
,[
line and staan line header it @
c.
high steam line flow in two out of three steam lines (one out of P.' Y Y two per line) in coincidence with either low reactor coolant J
id system average temperatura (two out of three loops) or low steam f.$N line pressure (two out of three lines)
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d.
two out of three high containment pressure signals.
Ni.' j
[l.~:'l o the overpower reactor trips (nuclear flux and differential r f. * '
temperature) and the reactor trip occurring upon actuation of the 4 T.g safety injection system.
l l -
o Redundant isolation of the MHf lines. A safety injection signal will y.4 close all m control valves, trip the main m pumps, and close the 7;g m block valves (safety grade). In addition, normal control action
- i...
will close the m control valves.
i.f@
mmh o Trip of the fast-acting safety-grade steam line isolation valves fg.'j$h (designed to close in less than 5 seconds with no flow) on:
, yyy h.
I a.
high steam flow in two out of three steam lines in coincidence E'j with either low reactor coolant system average temperature or low i
' J steam line pressure I
b.
two (two out of three) high-high containment pressure signals.
.ty.
Each steam line has a fast-closing stop valve with downstream check
!. ($
a.
' p.-;,
valve. These six valves prevent blowdown of more than one-steam generator for pA J,..
any break location even if one valve fails to close. Per breaks upstream of k
the stop valve in one line, closure of either the check valve in that line or the stop valves in the other lines will prevent blowdewn of the other steam
-II generators. Ibr all breaks, this arrangement precludes blowdown of more than f; X' cne steam generator inside the containment.
,p the m system for the plant includes a single turbine-driven pump (600
. h[r gpm) and two motor-driven pumps (300 gym each), each of which can supply all by.
three steam generators.
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[:,{k The AN flow from one motor-driven pump supplying a steam generator will i
$[dpiiie ensure that the heat removal capacity will exceed the minimum level required F *. */, a ;') 8 for decay heat removal after a MSIA.
.[jh The following signals are used for automatic initiation of the AM system Motor-driven pumos w
o Inw-Iow steam generator level (two out of three channels on any h
'&NN -
cp %
o Trip of both main feedwater pumps 65
.n*
?M,
o Ioss of all ac power mf kQi Nd
)
o safety injection V,s k
'Mrbine-driven pump 1
?(if,'
o Iow-Tow steam generator level (two out of three channels on two
+*
steam generators) 4 1,
f o Undervoltage on 4kV buses 1 and 4 (one out of two channels per bus).
The above systems are designed to meet safety-grade and IEEE Std 279-1971 y;
requirements.
.O 4,j The environmental qualification of safety-related electrical and h mechanical components is being reviewed separately by the KRC and is not
,.G a
- within the scope of this review.
i [
Y The Licensee's analysis determined that for the worst-case MSIA, which included runout AN flow for 10 minutes, the resultant containment pressure h
attained was 34.4 psig, compared to the design pressure of 42 psig.
y.,
q 1
3 p
Sufficient indications and alarms are available to the operator to determine that a MSIa has occurreds once this determination has been made, the 4
4 operator has aistinal actions to perform in order to isolate AN flow to the Q
ruptured steam generator.
It is conservative to assume that the operator will f
complete all required actions within the 10 minutes assumed in the analysis.
The review did not determino if the instrumentation that the operator fjgj relies upon to follow the accident and isolate the affected steam generator W%
we l Y iA f.
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conforms with the criteria in ANS/ ANSI-4.5-1980 [9] and Regulatory Guide 1.97
" [10].
e j
Since cavitaton of the AMt pumps is not expected to occur at runout' 7
,h,.g.,, conditions,nodamagewouldbeexpected.
("ih
.$ih 4
3.1.3 Conclusion h e Licensee's responsed [3 and 4] and the H. B. Robinson updated FSAR g
. [5] adequately address the concerns of Item 1 of IE Bulletin 80-04.
Se f/yy containment pressure response analysis and the design of the mitigating e
%3 systems satisfy the NBC's acceptance criteria. Regarding Item 1, it is j';
concluded that there is no potential for containment overpressurization p
, resulting from a MSLB with continued feedwater addition.
In addition, since 3
j the Ant pumps will not experience cavitation at runout flow conditions, the
[. J pumps will be able to carry out their intended function without damage.
3.2 REVIBt OF REACTIVITY INCREASE ANALYSIS 4
1
{
2e requirement from IE Bulletin 80-04, Item 2, is'as follows:
h
" Review your analysis of the reactivity increase which results from a main steam line break inside or outside containment. Bis raview should consider the reactor cooldown rate and the potential for the reactor to 5$j!
return-to-power with the most reactive control rod in the fully withdrawn Cj position. If your previous analysis did not consider all potential water f jij sources (such as those listed in 1 above) and if the reactivity increase G,, Q is greater than previous analysis indicated the report of this review 7'(
should include:
3 j
a.
Se 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 A
on the reactor system cooling, etc.,
.n.X f
b.
Se most restrictive single active failure in the safety g
injection system and the effect of that failure on delaying the
+
,_}j delivery of high concentration boric acid solution to the reactor coolant system,
[h]
?jgt; Se effect of extended water supply to the affected steam generator c.
g,tj.,l on r.he core criticality and return-to-power,
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He hot channel factors corresponding to the most reactive rod in the a
kr.
fully withdrawn position at the end of ' life, and the Minimum irg..
Departure from Itacleate Boiling Ratio (MDIER) values for the analysed transient."
.W:
3.2.1 Summary of Licensee Statements and Conclusions l
3, In regard to the reactivity increase resulting from a MSta with continued 9 feedwater addition, the Licensee stated [3]:
"Me worst case steam line break is assumed to occur at hot sero power T;
condition outside containment with offsite power available. At this y
time, the steam generator secondary side water inventory is at a maximum, Q
prolonging the duration and increasing the magnitude of the primary loop cooldown. With negative moderator temperature coefficient, this causes 4
reactivity insertion into the core. For conservatism, the most reactive control rod is assumed to be stuck out of the core when evaluating the shutdown capability.
With respect to additions of feedwater to the steam generator, main feed-
.j water flow at hot zero power when the accident initiates is approximately g
100-150 gpsm/ steam generator. Main feedwater isolates after approximately g
10 seconds, so main feedwater flow additions to the steam generator inventory are insignificant. Upon safety injection actuation, auxiliary
- u feedwater flow is initiated.
It is estimated that this flow would be g
established at approximately t + 40 seconds. At t + 38 seconds, safety
?J'.f injection has reached the core, and the cooldown reactivity transient has
- ~'
~
peaked and core Power is declining. S e auxiliary feedwater flow will not be sufficient to reverse this trend.
3
)
In summary, the core cooldown transient is driven by the blowdown of the full-steam generator. Continued small flow additions represented by
}
auxiliary flow capability are not significant contributions to the
, a reactivity transient."
t I
i y
Q 3.2.2 Evaluation j
i The Licensee's analysis of the core reactivity increase resulting from a
)
?
{MStawithcontinuedfeedwateradditionwasreviewedinordertoevaluate whether the following acceptance criteria were met e
fM 9 o Criterion 1.c - Ability to detect and isolate the damaged steam
- gJ.
generator y/' [q..
4 o Criterion 1.d - Changes in core reactivity increase 7?M,5I NM o criterion 3 Analysis assumptions.
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The FSAR [5] analysis of the reactivity increase resulting from a M8LB n2-and moferences 3 and 4 were reviewed. From that review, it was determined
.h;@ ;\\
',7G.fj that the analysis is conservative in its assumptions and that the assumptions i
L M,. "M are in accordance with those in Acceptance Criterion 3.
f M&U yygy In the worst-case MSta, which assumes no load conditions, a double-ended rupture at the steam generator exit, with offsite power available, the core JQ.31 returns to power at 14 seconds, a maximum core power of 454 is predicted at 38 (Q,#m Y'%, Aj seconds.
Shortly thereafter, 20,000 ppe boron solution reaches the core, rapidly shutting down the reactor. 1he calculated return-to-power did not
.ph lQ, i"
result in a violation of the specified acceptable fuel design limits.
ypg "I
Y AFW flow is initiated at approximately 40 seconds, since the : 0,000 ppa
.fb a e,y,t $j boron reaches the core before AFW reaches the steam generators, and since the 99 negative reactivity inserted by the boron significantly exceeds the positive h2 reactivity inserted by the cooldown caused by the addition of AFW, the core peak power will not be affected.
LA glh.j In addition, it can be assumed that the core transient is insensitive to
.u p runout AFW flow for the following reasons:
My,9 l
- q o early in the transient, the primary to secondary heat transfer rate
.,.,,, J (from the blowdown of the initial steam generator mass) is several l,$d orders of magnitude greater than that contributed by the additional
- .?.. %.f.h AFW flow due to runout
, yg<g
{;7.h o later in the transient (when the majority of the initial mass has i
l pf.C blown down), AFW flow becomes a dominant factor in determining the WU, magnitude and duration of the transient L
l hi o the limiting core conditions will occur within the first minute due to 4g 41.
the initial high cooldown rate cor.tributing to the reactivity addition 3
g<
which is terminated by the introduction of boron into the core region.
[.{f '
ge p,
since the limiting core conditions occur before the AFW flow becomes a
$.dmajor contributing factor, it can be concluded that the core transient is 4M g.,. j insensitive to the contribution of AFW flow, and therefore the assumptions of
'?
- the FSAR analysis remain valid.
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,3.2.3 Conclusion l
Se Licensee's response and FSAR adequately address the concerns of Item
- .{. ; 2 of IE Bulletin 80-04. All potential sources of water were identified, and Y:'.", '
,v..cl(although a reactor return-to-power is predict 9d, there is no violation of the g'lspecifiedacceptablefeeldesignlimits. Herefore, the FSAR analysis [5] of 3.w - the reactivity increase resulting from a MSLB remains valid.
- -)xv
[ys:.,yf13.3 REVIElf OF CORRECTIVE ACTIONS c.7:.I MD 'l The requirement from IE Bulletin 80-04, I' tem 3, is as follows:
I f(q?;,J s
- Q,'? !
"If the potential for containment overpressure exists or the reactor
~
jf;.c; return-to-power response worsens, provide a proposed corrective action t
y-and a schedule for completion of the corrective action.
If the unit is
- i?j;k' operating, provide a description of any interim action that will be taken f'T/
until the proposed corrective action is completed." -
! PrM.%
cc - :1 4
g a].3.3.1 Susmary of Licensee Statements and Conclusions
- .,d,3 '
-e
/,k h e Licensee stated (3):
h %,1 "As discussed above, no potential for containment overpressurization
. r, ],.
?((di:j 1.
exists, and the return to power response is very insensitive to r.fie addition of auxiliary feedwater. H erefore, no corrective action is
?,y y required."
h.
'd;5 /.
b5 3.3.2 Evaluation and conclusion
~.3.
j,'n%
ne Licensee's analysis determined t' st neither a containment overpres-
.d ', Q..
{
} surization nor a reactor return-to-power cith a violation of the specified g;]acceptablefueldesignlimitswouldoccurfromaMSLB. B erefore, it was concluded that no further action regarding IE Bulletin 80-04 is required of a [.[Y CP&L for the H. B. Robinson Steam Electric Plant Unit 2.
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CONCLUSIONS 2
1 s;.Njl8
[/
y~3p[1 With respect to the H. B. Robinson Unit 2, the ccnclusions regarding p
.r,
.h(fg Carolina Power and Light Company's response to IE Bulletin 80-04 are as r
j ust,
gy-follows:
p, dt.*lgi
'JfN - C o There is no potential for containment overpressurization resulting
,$$(OJ from a MSLB with continued feedwater addition.
WSTl
!IsTfi o the APW pumps will not experience cavitation at runout flow and 7 ll.
therefore can be expected to carry out their intended function during
$l};{,
the MSLB event.
f e
i o All potential water sources were identified and, although a reactor
.M return-to-power is predicted, there is no violation of the specified gy/ f(
i
[f increase analysis remains valid.
~ Ry;' f L-
- .yyl o No further action regarding IE Bulletin 80-04 is required.
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REFERENCES
.-.w.;
h.IN 1.
" Analysis of a PWR Main Steam Line Break with Continued Feedwater
- A Addition" h$i.h,4 Y
NRC Office of Inspection and Enforcement, February 8, 1980 g
[.
2.
"Overpressurization of the Containment of a PWR Plant af ter a Main
,j. %y y,
Line Steam Break" NRC Office of Inspection and Enforcement, October 1,1979
,(,
IE Information Notice 79-24 3.
L. W. Fury (CPriL) d[.,yf' c' '
Intter to J. P. O'Reilly (NRC, Region II)
$c
Subject:
IE Bulletin 80-04, Main Steam Line Break with Continued
$h Feedwater
?.yty May 9, 1980
.;.w
...O
- "n, a 4.
Standard Review Plan, Section 4.2 i$ h
" Fuel System Design" hh fy,[
NRC, July 1981 cE!!;$1
,y %
- m)ln,
.jk},@J:
J.:
5.
H. B. Robinson Steam Electric Plant Unit 2 Updated Final Safety Analysis Report l k'NS.'
Carolina Power and Light Company,1982 l
WL.)
l M
6.
Technical Evaluation Report W 3:l "PWR Main Steam Line Break with Continued Feedwater
[fg N
Addition - Review of Acceptance Criteria" Franklin Research Center, November 17, 1981 8 6j TER-C5506-119
'edn,2 k.
h' Q 7.
" Criteria for Protection Systems for Nuclear Power Generating i
QMyy Stations" pfC Institute of Electrical and Electronics Engineers, New York, NY, 1971 sh%
,i h Q[j 'g.
8.
Standard Review Plan, Section 15.1.5 l
l
- ms
" Steam System Piping Failures Inside and Outside of Containment (PWR)*
Mf "
NBC, July 1981 U.4.-
p NUREG-0800 l
s-g p.~
[ g]
9.
" Criteria for Accident Monitoring Functions in Light-Water-Cooled 4
a.
Reactors"
- t$j American Nuclear Society, Hinsdale, IL, December 1980 ANS/ ANSI-4.5-1980
, L%m.
. ~.
.. 4
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M)M
[GE.d'.
4 dbd Franklin,m rr. arch C. enter
$$[f9 Rese l
4 m en.
,ma
&&.y:
l qw:;$
l 1
1 l
w,xyc.m..+ w:.ag.y ;m.ss.; y n;a v.,;.m;.e y.
... a n ; ; 4 " g
^
$ -- d,. +. :.f :". (.T.49.: *W "; " i :1.i
~... ?r6 '. -
- '.<'~.".*.o 6-
p )I@Q
.,..a.
_1.
- ~.
~
' )h 5f*
g.?
- 4y
%'d5Mj l}.?!,
hh.i TER-C5506-137 3
fu.el yag 4,'s 4
A
?-) 3M.
Plant and Environs Conditions During and Pcilowing an Accident" 10.
' Instrumentation for Light 4sater-cooled Itaclear Power Plants to Assess fg
,)
Rev. 2
- 2. -.
W NBC, December 1980
?
~
11.
" Single Failure Criteria for PWR Fluid Systems" American maclear Society, Rinsdale, IL, June 1976 ANS-51.7/N658-1976
%~
76Ej7l 12.
" Quality Group Classifications and Standards for water, Steam, and Ej,K Radioactive-Weste-Containing Components of Itaclear Power Plants"
((@Vj, R*V.
3 WRC, February 1976 Regulatory Guide 1.26
^;r
?/,
i 13.
" Interim Staff Position on Environmental Qualification of Y
l Safety-Related Electrical Equipment" Rev. 1 h;s, NRC, July 1981 E' 7,,3
,f.'4 NUREG-0588 jN e.
reg.-
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