ML20090B511
| ML20090B511 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 05/24/1982 |
| From: | Holahan G Office of Nuclear Reactor Regulation |
| To: | Lainas G Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML082380335 | List:
|
| References | |
| FOIA-91-106 NUDOCS 8206080064 | |
| Download: ML20090B511 (6) | |
Text
.
s
(
f> ',y N.44 4*
DISTRIBUTION
, CentraFFTres?
OPAB Reading Gibla han WAY 2 4 481 MEMOPJWOUH FVR:
G. C. Lainas, Assistant Director for Safety Assessment Division of Licensing THAU:
T. A. Ippolito, Chief Operating Reactors Assessment Branch Division of Licensing rROM:
G. Iblahan, Section Leader Systems Section Operating Reactors Assessment 3 ranch Division of Licensing
SUBJECT:
GENERIC REC 0ft4ENDATIONS BASED ON Tl!E REVIEW OF TIE JANUARY 25, 1982 STEAM GENERATOR TtSE RUPTURE EVENT AT GINNA The May 3,1982 menorandum from lbrold Denton calling for the development-of generic recorriendations requested that members of the Ginna Task Force
- be involved to the extent practical.
Since ! was the team leader respon-sible for i'eview of the Plant System Response, on the Ginna Task Force.
I en taking this opportunity to present my recommendations relative to the generic implications of the Ginna event. These 'recomendations are presented in the enclosure.
I hue divided my recommendations into three categories to differentiate among (those items which (1) support the need for continuing on-going programs, 2) support the need for modifications to on-going programs, or (3) support the need for new generic programs.
As requested in the May 3,1982 menorardum on this subject. I have also identified these recornendations as relating to: Plant Systrns Response, Rnan Factors Consideratior.). Radiological Consequences, Organizational Response or Post-Event Activities.
G. Iblahan Section Leader Systems Section Operating Reactors Assessment Branch Division of Licensing
Enclosure:
- ,.-~N As stated
/
N h06g4pg
.p cc w/qn
.) re:
Seenyt.
A e
y' S.I..O..R.h..... d..L..U C.0..RABI o,,,co TI y,,_,,n],,7,,,, 7,7
,- [,,,
s....
suauwO
.9I.N.1 a,h,a,n,: d,n J
...S../.l. -.4... 8 2......
5./../.e... 8..2 m,,
hac ronu m oow Nacu om OFFICIAL RECORD COPi usom im-mwo
O 4
x.
f G. C. Lainas 2-cc w/ enclosures D. Eisentwt R. Mattson G. Iblahan J. 1.yons
- 0. Crutchfielti T. Spets B. Sheron W. Lanning W. Hfoners d
i orrect) sununc6
,D(Tt )
.......... +... * *. *
~
OFFiC1AL RECORD COPY iu.m mi n.
.-.........,....,.....a
ANCLOSURE i
y 4
l
. ecomendations supporting the Need for Continuing On-Goina Programs 1.
R 1.1 Plant System _s Response, 1.1.1 Rc' 'or vessel tevel Measurement T h.
asek of a reactor vessel level measurement system significantly comp 1f.-
cased the Ginna event. It was the presence of a steam bubble of unknown size in the reactor vessel upper head and the fear of increasing the size of that t,ubble that caused the reactor cperators to delay termination of high pressure safety injection. It uns the continuing safety injection which lead to
.ie overfilling of the steam generator and the opening of the Steam Generator safety valve. Insta11ction of a reliable reactor vessel level measurment systeai would significantly aid in nanaging SGTR events.
1.2 Jbman Factors Considerations 1.2.1 Peview M SGTR with teneurrent railure of Primary or Secondary Pelief or Ffetyvalves The Ginna event was an SGTR which included both primary and secondary systm valve failures. The PORY failure to close was quickly and effectively dealt with but the leakage of the Steam Generator Safety Valve went
. unnoticed and the complications it introduced in handling the event were not appreciated by the plant operators. Probluns of multiple failures.
beyond,,the design basis assumptions, are being handled through the TMI Action Plan item I.C.1.
This program requires operator training and emer-gency procedures for the more important and more likely of the possible multiple failure evnts. Ceoletion of Action Man item I.C.1 is an appropriate and suf ficient,pneric response to this concern.
1.3 Radiolo2 feel Consequenceg, No coments 1.4 Organizational Response.
No coments 1.5 PostEventActivlties i'
No coments f
2.
Recomendations Supporting Modifications to On-Goinn Programs 2.1 Plant Systeus Response _.
6
.,.__m.....
1
.2-2.1.1 Pressurtzed Themal Shock During the Gtnna SGTR event the isolation of the steam generator with the ruptured tube plus the tripping of the reactor coolant pumps resu t
The injection of cold ECCS water tnto the stagnant the other loop.
During loop resulted in a rapid decrease in the cold leg temperature. event the cold wate the January 25, 1982 never flowed into the reactor vessel creating a potential therinal shock Ibwever, the Gfnna event did identify an important phunomena which may not be receiving sufficient attention in the pressurized Thermal prob 1 m.
Shock program, that is, the influence of steam generator isclation and The tiow stagnation on thn potential for Reactor Yessel Thermal Shock.
action plan on U51 A-47 should be modified to specifically address and resolve this issue.
?,1.2 Reactor Coolant Pump Trip _ Requirer ent__
~
The Ginna This has been a difficult on-going issne for several years.
SGTR event is only the latest event to be compitcated by the requirement to trip the Reactor Coolant Pumps (RCP) on a suspected small break LOCA.
I am fully aware that the suggestion to trip the P"P's is a Westinghouse-However, it appears developed recommendation which the staf f accepted.
clqr that all of the compitcations of the January 25, 1982 event probably woNN have been avoided if the RCP's had not been tripped.
Since itcensing credit for rapid manual action to trip the RCP's is not consistent with past Itcensing practice, it does not appear that manual RCP trip resolves the legal 10 CFR 50.46, Appendix K concerns; and in tems l
of safety significance, l~ be,1teve that allowing continued RCP operations I
is desirable for the follo' wing reasons. First, the size, location and There-timing of small 1.0CA's which would exceed 22000F !s very limited.,Second, the benefits fore, such events would be expected to be quite rate, of RCP operation in terms of heet removal capabtitty, plant control and Third, if.a small 1.0CA did plant-transtent understandablitty are grer begin to lead to unacceptable consequences because of excessive inventory loss associated with RCP operation, the existing, approved procedures for responding to indications of inadequate core cooling (wculd lead theprfary and sec plant operators to take the necessary corrective action depressurization). Therefore, while long tem resolution of this is 2.1.3 Steam Generator Overf111 The Gtnna event resulted in an overf t11t_ng of the steam generator and a flooding of the main steam line up to the H51Y. Overfilling of PWR SG has also occurred in the past yet there continues to be considerable confusion AEOD has ratsed relative to the requirment to analyse such occurrences,
s a
. this issue on several occasions and has stated that the main steam linea program is is not designed for the loads associated with such flooding.
needed to review and document the full spectrum of concerns and rsquire-ments in this area and to determine the degree of compliance in operating This could be done in the frame work of the recently begun U31 plants.
in this area.
2.2 Human Factors Consideration _s
.1.1 Accident Monitoring __
The instrumentation used to monitor the course of the January 25,1982 event had several deficiencies including non-redundant monitoring of the RCS pressure, failure of the position recording for secondary reitef and safety valves and no flow or valve position monitoring on RCS leakage path such as the letdown relief valve and the seal-return line relief valve, Implementation o/ Regulatory Guide 1.97 on operating reactors would resolve p.'ob1ms associated with monitoring important parameters such as RCS pressure, however, the guide may need to be modified to more fully address the monitoring of primary and secondary leakage in this area.
2.2.2 ; Emergency procedure Reviews, D ring the Ginna event the formation of a steam bubble in the reactor ves Upon upper head occurred but had not been expected by the plant operators.
annlysis of the event it is clear that steem fomation should h NSSS vendors and the lictasee,s to, perform best-estimate, plant spe:ific expected.
The review process analysis on the developmen, of, emergency procedures.
t for approving plant emergency procedures and guidelines should be modi to require plant specific analysis in the development or at least in the ve.rification of emergency procedures.
2.2.3 Shif t Technical Advisor (STAl
,cocess of During the Ginna event the STA involved himself directly in thoperators.
handling the event by reading the mergency procedure to the i.a This is clearly not the independent, thoughtful. " stand back' ovr role originally intended.
This subject needs,
original STA concept is not being properly implemented.to b and strengthen the requirment or to revise the present program.
2.3 Radiological Consequences No comments 2.4 Organizational o?sponse.
No coments
4
.4.
2.5 Post Event Activities No conments 3.
Recommendations for New Generic Programs The extensive reliance on "non-safety related equipnent" during the Ginna event indicates that the FSAR and SRP reviews of safety related equipment has been done with much too narreu a view as to what equipment is consider ed. Regulatory Gaide 1.97 will resolve many ef the issues related to instrumentation but a program to identify truely safety related equipment based on operating experience and energency procedure review This information should be reflected in SRP modifications where is needed.
Decisions on operating plants should be made af ter the scope necessary.
of the problem is better understood.
3.2 lbman Factors Considerations __
No comments 3.3 Radioloqical Consequences 3.4 Oganizational Responst Comments 3.4.1 ' Extensive training is needed for en effective NRC incident response.
on this matter were developed and documented in a March 4,1982 memorandum from W. Binners to H. R. Denton and R. C. DeYoung, " Improving the NRC Incident Response Technical Assessment Capability." My comments on this subject are included in that document.
3.5 Post Event Activities The finding of loose pirts and the associated damage in the Ginna Steam 3.5.1 Ganerator and the prior history of undetected loose parts damage clearly indicate the need for periodic she11 side, visual inspection of steam gener-Requiring such inspection at the presently planned outages for ators.
eddy-current testir;g appears to be a reasonable approach.
4
i i,
5', 4.
d i
~
DISTRIBUT10N (CentraT~tTl'e~s !
ORAB' Read (ng Gibla han MAY 2 41992 MEMORANDUM FOR:
G. C. Lainas, Assistant Director for Safety Assessment Division of Licensing THRU:
T. A. Ippolito, Chief Operating Reactors Assessment Branch Division of Licensing FROM:
G. Iblahan, Section Leadu Systems Section Operating Reactors Assessment Branch Division of Licensing
SUBJECT:
GENERIC RECO?t4ENDAT10NS BASED ON THE REVIEW OF THE JANUARY 25, 1982 STEAM GENERATOR TUBE RUPTURE EVENT AT GINNA The May 3,1982 menorandum from Harold Denton calling for the developnent of generic recomendations requested that members of the Ginna Task Force be involved to the extent practical. Since I was the team leader respon-sible for review of the Plant System Response, on the Ginns Task Force.
I am taking this opportunity to present my recomendations relative to the generic implications of the Ginna event. These 'rtcomendations are presented in the enclosure.
I have divided ry recommendations into three categories to differentiate among those items which (1) support the need for continuing on-going programs, (2) supWrt the need for modifications to r.a-going programs, or (3) support the need for new generic programs.
As requested in the May 3,1982 memorandtn on this subject, I have also idantified these recomendations as relating to: Plant Systems Response, iknan Factors Considerations, Radfological Consequences Organizational Response or Post-Event Activities.
G. iblaban, Section leader Systens Section Operating Reactors Assessment Branch Division of Licensing
,7/ktfDTMDtg1,
Enclosure:
As stated
-8206
(
~ 080064 829094-==-ooooop. sop g.
cc w/y,;
j r e *-
See n
,e S.L..O..P.h.......
3 C... 0..RAB : d(.. y......
...................,w x.{'M w.m
. 9"o.l a.h,a,m,d n, TI py,,,,, g,,
,,( / Q.... -.,.,pg/;,
.. 5 /.2...h... 8. 2.....
. 5/.).h... 8. 2....
m, >
nreir Ai pr CORD COPY uscoo. an-w
h 2-G. C. Lainas cc w/ enclosure D. Eisenhut R. Mattson G, Holahan J. Lyons
- 0. Crutchfield T. Speis B. Sheron W. Lanning W. Hinners M
9 e
orier>
$URNAME)
/
s
.............l ahp. )........:...............
e t' L? 4 e 4 A I L'/ C ( A Q f3 rAAV
.,r.,,,
1 "HCLOSURE l
s l
1.
Recomendations Sepporting t_he Need for Continuing On Going Programs 1.1 Plant Systms_ Response 1.1.1 Reactor.ssel Level Measurement The lack of a reactor vessel icvel measurement systm significantly comp 11 cated the Ginna event, it was the presence of a steam bubble of unknown size in the reactor vessel upper head and the fear of increasing the size of that bubble that caused the reactor operators to delay temination of high pressure safety injection. It was the continuing safety injection whi:n lead to the overfilling of the steam generator and the opening of the Steam Generator safety valve. Installation of a reliable reactor vessel level measurement system would significantly aid in managing SGTR events.
1.2 lbman Factors Considerations 1.2.1 Pevis c3 574 with _concurret railure of Primary or secondary Pelief _ or Mtety valves The Ginna event was an SGTR which included both primary and secondary systs valve failures. The PORV failure to close was quickly and effectively dealt with but the leakage of the Steam Generator Safety Valve went unnoticed and the complications it introduced in handling the event were
~
not appreciated by the plant operators.
Problems of multiple failures, Deyond the design basis assumptions, are being handled through the TMI Action" Plan item 1.C.1.
This program requires operator training and ener-gency procedures for the more important and more likely of the possible rT1tiple f ailure events.
Ceoletion of Action Plan item I.C.1 is an appropriate and suf ficient,;tneric reponse to this concern.
1.3 Radiological Consequences No coments 1.4 Organizational Response No coments 1.5 Post Event Activities, L,,
No coments 2.
Recommendations Supporting Modifications to On. Going. Programs 2.1 Plant Systems Rnponse
--~--"--~-~-..-___._m..m_
~
~
t
.2 2.1.1 Pressurtzed Thema_1 Shock During the Gtnru SGTR event the isolation of the steam generator with the ruptured tube plus the tripping of the reactor coolant pumps resulted in natural circulation flow in one loop and near stagnant conditions in The injection of cold ECCS water into the stagnant the other loop.
During loop resulted in a rapid decrease in the cold leg tmperature. event the cold water in t the January 25, 1982 never flowed into the reactor vessel creating a potential thermal shock ibwever, the Gtnna event did identify an important phenomena which may not be receiving suf ficient attention in the Pressurized Themal probl em.
Shock program, that is, the influence of steam generator isolat, ion and The flow stagnation on the potential for Reactor Yessel Thermal Shock.
action plan on US! A 47 should be modified to specifically address and resolve this issue.
2.1.2 Reactor Coolant Pump Trip Requirement The Ginna This has been a difficult on-going issue for several years, SGTR event is only the latest event to be complicated by the requirement to trip the Reactor Coolant Pumps (RCP) on a suspected small break LOCA.
I am fully aware that the suggtstion to trip the R:P's is a Westinghouse-Fbwever, it eppears developed recommendation which the staff accepted.
clear that all of the compiteations of the January 25, 1982 event probably would have been avoided if the RCP's had not been tripped.
1 Since Itcensing credit for rapid manual -tion to trip the RCP's is not consistent with past ifcensing practice, it does not appear that manual RCP trip resolves the legal 10 CFR 50.46, Appendix K concerns; anc in tems of safety significance, I be,lieve that allowing continued RCP operations is desirable for the follo' wing reasons. First, the size, location and There-timing of small LOCA's which would exceed 22000F is very limited.Second, the benefits fore, such events would be expected to be quite rate.
of RCP operation in tems of heat removal c;.pabtitty, plant control and Third, if a small LOCA did plant-transient understandability are great.
begin to lead to unacceptable consequences because of excessive inventory loss associated w'th RCP operation, the existing, approved procedures for responding to indications of inadequate core cooling (would lead theprimary and sec plant operators to take the necessary corrective action depressurization). Therefore, while long tem resolution of this tssue is continuing, the NRC interim position requiring RCP trip stould be changed. '
2.1. 3 Steam Generator Overfill The Gir.na event resulted in an overf t11tng of the stean generator and a Overf t111ng of PWR SG has flooding of the main steam line up to the NSIV.
also occurred in the past yet there continues to be considerable confusion AE00 has raised relative to the requirtnent to analyse such occurrences.
d
^^
A i
3 this issue on several occasions and has stated that the main steam linea program is is not designed for the loads associated with such flooding.needed ments in this area and to determine the degree of compliance in operating This could te done in the frame work of the recently begun US!
pl ants.
in this area.
2.2 Human Factors Considerations
.2.1 Accident Konitoring_
25, 1982 The instrumentation used to monitor the course of the January event had several deficiencies including non-redundant mon safety valves and no flow or valve tosition monitoring on RCS leakage path such as the letdown relief valve and the seal-return line re resolve problems associated with monitoring important parameters such as valv e.
RCS pressure, however, the guide may need to be modified to more fully address the monitoring of primary and secondary leakage during in this area.
,2.2.2 Emergency Procedure Reviews During the Ginna event the formation of a steam bubble in the reactor vessel Upon upper head occurred but had not been. expected by the plant operators.
analyhis of the event it is clear that steam formation should have be NSSS vendors and the licensee,s to, perform best-estimate, plant spe:ific expected.
The review process analysis on the developmen; of emergency procedures.
for approving plant emergency procedures and guidelines should be modified to require plant specific analysis in the development or at least in the verification of emergency procedures.
2.2.3 _Shif t Technical Advisor (STA)
During the Ginna event the STA involved himself directly in the process of handling the event by reading the energency procedure to the plant operators.
This is clearly not the independent, thoughtful
" stand back" overviewIt appears from this and other role originally intended.
This subject reeds original STA concept is not being properly implenented.to be carefully and strengthen the requirement or to revise the present program.
2.3 Radiological Consequences No comments 2.4 0_rganizational Response No comments l
u
i 1
.4.
Post jye_nt Activities 2.5 o
No car.ments 3.
Recomendations for New Generic _ Programs _
The extensive reliance on "non-safety related equipment" during the Ginna event indicates that the FSAR and SRF reviews of safety related equipment has been done with much too narrow a view as to what equipment Regulatory Guide 1.97 will resolve many of the issues related to instrumentation but a program to identify truely safety related is considered.
equipment based on operating experience and emergency procedure is needed.
Decisions on operating plants should be made af ter the scope necessary.
of the probitsn is better understood.
3.2 Jbman Factors Considerations No coments 3.3 Radiological Consequeneet 3.4 _Organizatiomi Response, Coments Extensive training is needed for an effective NRC incident response.
on this matter were developed and docunented in a March 4,1982 memoraridum 3.4.1 from W. Minners to H. R. Denton and R. C. DeYoung, " Improving the NRCMy comen Incident Response Technical Assessment Capability."
sub, ject are included in that document.
3.5 Post Event Activities
.~
The finding of loose parts and the associated damage in the Ginna Stesm Generator and the prior history of undetected loose parts damage clearly 3.5.1 indicate the need for periodic she11 side, visual inspection of steam gener-Requiring such inspection at the presently planned outages for ators.
eddy-current testing appears to be a reasonable approach.
_ _ _ _ _ _. _. _... _. _ _ _