ML19339A216
| ML19339A216 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 09/24/1980 |
| From: | Dunn C DUQUESNE LIGHT CO. |
| To: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| IEB-80-18, NUDOCS 8011030288 | |
| Download: ML19339A216 (10) | |
Text
79e_
o u
4.
'Af M
om 4se-sooo 43$ Seath Avenue Pfttsourgh. Pa.
September 24, 1980 United States Nuclear Regulatory Commission Office of Inspection and Enforcement Attn: Boyce H. Grier, Regional Director Region I 631 Park Avenue King of Prussia, Pennsylvania 19406
Reference:
Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 I.E. Bulletin No. 80-18 Gentlemen:
We have reviewed I.E. Bulletin No. 80-18, " Maintenance of Adequate Minimum Flow Thru Centrifugal Charging Pumps Following Secondary High Energy Line Rupture," as well as the Westinghouse "Part 21" letter on the subject.
Using the calculation method provided by Westinghouse and plant data from Beaver Valley Unit 1, a calculation was performed to determine whether minimum pump flow is maintained during parallel safety injection operation of two centrifugal charging pumps (CCPs).
The results of our calculations indicate that the availability of minimum cooling flow for the CCPs is not assured for all conditions.
Licensee Event Report 80-060 was submitted identifying this potential condition.
a.
Since a plant-specific concern was identified, the modifications recommended by Westinghouse have been reviewed for incorporation.
As a result:
1.
The removal of the safety injection initiation automatic closure signal from the CCP miniflow isolation valves has been accomplished.
2.
The Onsite Safety Committee (OSC) has approved revisions l
to the Emergency Operating Procedures to instruct the l
operator when to close or reopen the CCP miniflow isolation I
valves.
l 1
1 8011 030 ggy
1 Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 I.E. Bulletin No. 80-18 Page 2 b.
In order to evaluate the impact of implementing the recommended interim modifications, the OSC reviewed the Westinghouse evaluation of the sensitivity of the FSAR transient analysis to the emergency 1
operating procedure interim modifications.
These studies show that the accidents evaluated are relatively insensitive to the recommended modifications.
Further, the accidents evaluated will give results that satisfy acceptance criteria as long as the CCP miniflow is isolated within 10 minutes of event initiation. However, small LOCA sensitivity studies with one SI train operating confirm that small LOCA analyses require miniflow isolation within 10 minutes.
To comply with the recommended modifications, the operator can isolate miniffow at any point in the depressurization transient prior to RCS pressure reaching the RCP trip setpoint.
The CCP a
miniflow can be isolated by the operator in the Control Room.
Should a repressurization transient occur, the operator can open CCP miniflow at any point between the RCP trip setpoint and 2000 psig.
Such operator actions will ensure that plant accidents satisfy acceptance criteria and protect the CCPs from consequential damage during the repressurization transient that accompanies a i
secondary system high energy line rupture at high initial power levels.
c.
A review of the Operating Manual shows that isolation of CCP miniflow can be accomplished with either train available and can be accomplished without offsite power available.
d.
Based upon the results of these studies, the flow available from the CCPs, as modified, is sufficient to justify continued appli-l ca'sility of any safety-related analysis which takes credit for i
flow from these pumps.
Since the safety analysis for LOCA and HELB remain valid, the e.
Technical Specifications derived from these analysis remain uneffected.
A copy of the Westinghouse evaluation is attached.
The letter which transmitted I.E.Bulletin 80-18 requested that we provide an estimate of the manpower expended as a result of this Bulletin.
We estimate that 50 man hours have been expended in conducting reviews required by this Bulletin and preparation of the report. Manpower expended i
for corrective actions to date is 16 man hours.
1
Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 I.E. Bulletin No. 80-18 Page 3 If you have any questions concerning this response, please contact my office.
Very truly yours, i
C. N. Dunn Vice President, Operations Attachment ec:
Mr. D. A. Beckman, Resident Inspector U.S. Nuclear Regulatory Commission Beaver Valley Power Station Shippingport, Pennsylvania 15077 U.S. Nuclear Regulatory Commission c/o Document Management Branch Washington, D.C.
20555 U.S. htclear Regulatory Commission Director, Division of Reactor Operations Inspection Office of Inspection and Enforcement Washing' ton, D.C.
20555
...l
.N l.
.n w
2
' Westinghouse Water Reactor d
~M'8 8=8 N5=
0
,[,] sa2r3 Sectric Corporation Olvisicas
,,,, Pir:scurp Pennsylvansa 15220
~ June'l8, 1980 DLW-80-68 Ref:
DLW-80-50 May 8,1980
~~
Mr. J. A. Werling, Plant Superintendent Beaver Valley Power Station Duquesne Light Corpany P. O. Box 4 Shippingport, Per.nsylvania 15077
Dear Mr. Werling:
Duquesne Light Company Beaver Valley Unit No.1 CENTRIFUGAL CHARGING PUMP OPERATION FOLLOWING SECONDARY SIDE HIGH ENERGY LINE RUPTURE AND SMALL LOSS OF COOLANT ACCIDENTS The referenced letter previously transmitted information concerning the potential for consequential damage of one or more centrifugai charging pumps (CCP) in the evet of a secondary cystem high energy line rupture.
The previous information (Reference 1 - W letter NS-TMA-2245 dated May 8,1980 to V. Stallo, NRC, which was'" attached to the above refer-enced letter) included a calculational method and a sample ' calculation to permit evaluation of this concern on a specific plant basis.
Should a plant specific problem be identified, several recommendations were provided for the interim until necessary design modifications could be implemented to resolve the problem. supplements the previous information and provides a generic evaluation of the sensitivity of the FSAR transient analyses to the emer-gency operating procedure Interim modifications proposed in Reference 1.
This infomation is provided to permit plants which identify a plant specific concern to evaluate the impact of implementing the recommended interim modifications.
., y.
_L -
t 4=.
o
.~.
.. l Page 2 June 18, 1980
.g g
It is anticipated that this type of information will be requested by the NRC of any licensee that confirms a CCP operation problem and 'ccmmits to implementing one of the Westinghouse reccmmended interim modifications.
- 2;a Very truly yours, J
,_M"- -
[ F. Noon, Mana'ger Eastern Region & WNI Support SR3/AA10 Attachment
' -- W -
cc:
G. W. Moore R. E. Martin
~ ~
N. P. Williams J. J. Carey J. D. Sieber
"~
F. W. Knowles W f._.,-
. zl e
W
.,n 1
m.
.e.
e
.e 6
~.. -
gG e
h e
e e
n 9.e% -
l
(
i
. Attachmeni 1
. 7.
-D D
- D O-ow m.
A
-- CENTRIFUGAL CHARG1NG PUMP. OPERATION.
e63 e'
-kijr - FOLLCWING SECONDARY SIDE HIGH ENERGY LINE RUPTURE
..w.
~..
- 5...
..: ~. _ -
. m:.
. ~.
.. E L Reference 1:- NS-TMi-2245, 5/8/80
~
Reference 1 notified the NRC of a c:ncern for c:nsequential damage of cne cr mcrs centrifugal charging pu=ps (CCF) follcwing a sec:ndary system
~
high energy ifne rupture.
Reference 1 included a calculaticnal mathed and sample calculatien' to per::it evaluatien of this c:ncern en a plant specific basis.
Should a plant specific problem be identified, 'destinghcuse provided several' rec:mmendations for the interim until necessary design modificatiens can be implemented to resolve the problem.
These rec:m enda-tiens included two proposed interim mcdifications which included:
1.
Remove the safety injection initiatien aut:=atic closure signal feca the CCP minificw isolaticn valves.
2.
Mcdify p.lant emergency operating precedures to instruct the acerator :::
a.
Cicse the CC? minificw isciatien valves when the ac ua! RC5 press.'re drops to the calculatad pressure for manual reace:r c:cla.<' pump trip.
b.
Recpe the CC? minificw isciaticn valves should the wide range RCS pressure subsacuently rise to greater than 2000 psig.
Prior to making this recemendation, 'destinghousa evaluated the im:act of the rec: :: ended operating precedure mcdifications en the results of the varicus accidents which initiate safety injection and are sensitive to CCF flow delivery.
The accidents evaluated in detail include secondary system ruptures and the spectrum of small loss of c:clant accidents.
The analytical results for steam generator tube rupture and large loss of c:clant ac:fdent are not sensitive to a reduction in CCF ficw of the magnitude that resules frem the rec:m ended mcdificaticns.
This letter functicns to supplement l
Reference 1 and identify the sensitivity of the accident analyses to the rec:m, ended mcdificaticns.
This evaluaticn is generic in nature.
y
-2 D * *AD*A'T W k
Secondary Svstem Ruoture Sensitivity analyses have been perfomed for sec:ndarj-high energy line ruptures t: avaluata the impact of reduced safety injection ficw due t:
nomally open miniflew isolation valves.
These analyses indicate an insignificant effect on the plant transient respense.
A.
Feedline. Rupture Following a feedline rupture, the react:r c:oiant pressure will reach the pressuri:ar safety valve setpoint within approximataly 100 rac:nds assuming maximum safeguards with the pcwer;cperatad relief valves inoperable. Mith minimum safeguards, the react:r c:alant pressure will not reach the pressurizer safety valve set;oint untti approximataly 300 sec:nds.
The time that the react:r c:clant system pressure remains at the pressurizer safety valve set;cint is a function of the auxiliary feedwatar ficw injectad int: the ncn-faultad steam generat:rs and the,
time at which the operat:r is assu=ed :: take action.
With the mini-ficw isolation valves open, the peak react:r c:clant systam pressure,
and the water discharged via the pressuri:er safety valves are insignifi-cantly changed from the F5AR results.
3.
Staamline Rupture The effects of =aintaining the minificw isolation valves in a nor= ally open position was also investigated folicwing a main steamline rupture.
For the condition II " credible" steamline rupture, the results of the transient with the minificw val.ves cpen shewed that the licensing critation (no return t: criticality after react:r trip) centinues t:
be met.
Tne c:ndition III and IV main steamline ruptures were also reanaly:ed assuming the minificw valves were open.
The results of the analysis shewed that, even with reduced safety injection ficw into the core, no ONS cccurred for any rupture.
~
3 I
l Small less of Ccclant Accidents D
Sensitivity analyses have been perfemed to evaluate the imcact of reduced safety injectica flew on small break loss of ecolant accidents (LOCAs).
These analyses indicated that minificw isolation can be delayed,'but it must cccur at scme tima' ints the small break LOCA transient in crder to lim u the peak clad tamperature (FCT) penalty.
The prepcsed mcdification delays mi~nificw isolatien and rsduces SI ficw delivered by approximately 45 gpm at 1250 psia during the delay time period.
The impact of this modification was evaluated based en two isolation times:
- 1) The time equivalent to the RCP trip time, and 2) apercximataly 10 minutas in the transient, or just price to system drain to the break for the wors small break si:es.
The sac =nd time was av.aluated to detamine the impact if the operaccr does not isolata miniflew within de prepcsed prescribec time.
The spectrum of s=all break sizes are c:nsidered to anc:m: ass all possible s=all breaic scenarics.
Only ccid leg. break Iccatiens-are c:nsidered sines they will c:ntinue to be limiting in tarms of FCT.
Very br.all breaks tnat do not drain the RC5 er unc:ver the care, and A.
maintain RCS pressure above sac:ndary pressure ('< 4 " diametar).
For these break si:es, it is quita possible that the cperat:r may never isolata the minificw line, since the :;ressure setpoint will not be reached, and c:ntinued pumced SI degradation will persist.
Hewever, this will have nc adverse c:nsequences in terms of core uncovery and FCT.
No care unccvery will be expected for the degraded SI case, similarly to the base c:mparisen case with full SI.
The only effect would be a sligntly Icwer equilibr: tion pressure fer a give'n break si:e.
, S.
Small breaks that drain the RC5 and result in the maximum cladding temceratares (2" < diameter < 6").
This range of break si:es recresents. the worst small bieak si:e for
'^
4-Attachment i most plants as detar.nined utili:ing the currently c.pproved Oct:bar 1975 Evaluation Medal version, as shcwn in WCAF-8970-F-1.
If minificw is isolated at the RCP trip set;:cint rather than the *S signal, a reduc-tion in safety injection ficw of less than 45 gpm results, averaged for the app'reximately 50 sec:nd pericd of time separating the two events.
This reduction in' RC3'licuid inventary resu'lts in care uncovery less than one sec:nd earlier, and has a negligible impact on FCT.
If mini-flew is isolated at.the time of core uncovery, or approximataly 10 minutes for break si:ss in this range, a greater rsduction in RCS liquid inventerv esults in a core uncovery 10 sec:nds earlier in the transients resulting in less than a 10*F FCT penalty for the wors: si:e small break.
This would not result in any present FEAR smail break analysis bec: ming more limiting than the c:rrescending large break LOCA FSAR analysis.
If minificw isolaticn 'dcas not ec:ur at any time into the transient for this categ:ry of small LOCA, a FCT penalty of 200*F cr more c:uld ec:ur.
Small break sizes larger than t. e worst break through the intemediata 5
C.
break si:ss (> 5" diameter).
Sreak si:es in this range have been detemined to be non-limiting for small break util1 ing the currently approved October 1975 Evaluatien Medal, WCAF-8970-F-A.
If minificw isclatien ec:urs at the RCP trip time for these break sizes, the negligible effect cri PCT presen:ad above also applies.
Similarly, if isciation cc:urs prier to c:re sncovery, the small (<.10*F) FCT penalty will result as well.
- Hewever, I
for these larger break si:ss, the time of first core uncovery occurs prior to 10 minutes.
If minificw isolaticn. is not performed until 10 minutes, reduced SI will be delivered during the core unc:very time, which can,have a greater impact en FCT.
Studies indicate a potential PCT penalty of 40*F resulting for these ncn-1imiting break si:es if miniflew is not isolated until 10 minutes.
This is not expected Oc shift the worst break si:e to larger breaks, since these breaks are l
typically hundreds of degrees less than smaller limiting small breaks analyzed with the currently approved Evaluatien Medel.
l
--=
9
~.
5-For all F5AR small LOCA inalyses, cne c:mplete train failure is assumed.
It is clear that two charging pumps without minificw isolation provides mors flow than one pump with miniflew isciation.
The impact presented in this evaluatier$ maintains the ene train failure and assumes no minificw isc1 tion for the rdaining pumo.
If both pumps wers operating, the FCT.results would be much icwer than present F'5AR calculations even if minificw isola-tien is not assumed tc occur for the two pump case.
In this situstien, the plant F5AR small break calculations remain c:nservative.
These sensitivity studies fer= the basis for the rec:= ended intarim mcdificatiens to the emergency operating precadures.
The accidents evalu-ated are relatively insensitive tc de rec:cended mcdifications.
- Further, the ac:idents evaluated will give results that satisfy ac:actanca critaria as long as the CCF minificw is isolated within 10 minutas of event initiatien.
Hcwever, small LOCA sensitivity studies wi-h,cne SI train acerating c:nfirm that small LOCA analyses require minificw isolatien within 10 minutas.
To c mply with the rec:= ended mcdificaticns, the operator can isciate mini-ficw at any point in tne deprerturi:stien transient prior t: p,C5 pressure reaching the RCP trip set;cint.
Shculd a repressuri:ation transient cc:ur, the cperat:r can open CCF minificw at any point between the RC? trip set-point and 2000 psig.
Such operat:r acticns will ensure that plant accidents satisfy acceptance critaria and protect' the CC?s from consequential damage during the repressurization transient that ac::=panies a sec:ndary systam high energy line rupture at.high initial pcwer levels.
a6
.