ML20086K659
| ML20086K659 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/11/1983 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Beverly Clayton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20079G498 | List:
|
| References | |
| FOIA-83-243, FOIA-83-A-18 NUDOCS 8402060243 | |
| Download: ML20086K659 (12) | |
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'O UNITED STATES
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NUCLEAR REGULATORY COMMISSION
! Yi e iE WASHINGTON, D. c. 20555 3, 1,'.
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p 7) o y~'.'.. f April 11, 1983 Docket No. 50-280 MEMORANDUM FOR: Brent H. Clayton, Acting Chief Procedures and Systems Review Branch, DHFS FROM:
John F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing
SUBJECT:
REQUEST FOR TMI-1 STEAM GENERATOR TUBE RUPTURE PROCEDURES AND GUIDELINES REVIEW t
As a follow-up to discussions between Sam Bryan (PSRB) and Jim Van Vliet (ORBf4), we request that PSRB review the subject guidelines and procedures and provide us with a written Safety Evaluation input. We will then incorporate your input into the TMI-l steam generator repair Safety Evaluation. We request that you provide your input by no later than May 9,1983 by which date all the SER inputs are expected.- If it becomes obvious that you cannot meet the May 9 date, please advise us immediately.
We have requested that GPU submit the guidelines and procedures; we should receive them shortly. GPU has previously submitted a summary of the guidelines which is forwarded herewith.
W
,u johhF.Stol:, Chief
! Operating Reactors Branch #4 Wvision of Licensing
Enclosure:
GPU Topical Report 008, Rev. 2,Section X cc w/ enclosure:
SBryon cc w/o enclosure:
JC11fford DZiemann Glainas CMcCracken VBenaroya TMarsh BMann JVan Vliet 8402060243 831027
Contact:
J. Van Vliet 27380 D$0 SOB 3-A-19 PDR
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X.
OpERATIONA' CONSIDERATIONS The operational concerns of pri=ary to secondary leakage were evaluated.
Concerns inzluded leakage monitering during normal operations in both stea=ing and nonsteaming conditions, and sampling j
steps to be taken when leakage is detected.
In addition, a program j
has been for=ulated that includes procedure review and operator training which will provide improved operator guidelines for dealing with tube leakage and tube rupture events.
The operational guidelines discussed in this section are applicable during normal operation with lov levels of primary to secondary 1
leakage. A more detailed description of these guidelines can be found in reference 58.
For primary to secondary leakage ratio of 50 gpm or greater, these guidelines will be superseded by tube rupture guidelines as discussed in Section X.B.
s Operational concerns can be grouped into three general areas *
{
1.
Primary to Secondary leakage which includes leakage detection methods, and actions required based on primary to secondary leakage.
2.
Radiological concerns which include detection methods, worker protection ceasures and plant discharge limits.
3.
Secondary side chemistry limits ' based on boron and lithium con-centrations.
l This section summarizes the guidelines for operating with tube leakage.
n A.
primary to Secondary Leakage During normal power operation the methods which will be used to
[
detect and monitor leak ge are-'
i 1.-
Of fgas continuous monitor (RMA-5) i i
2.
Tritium samples _ from the condensate and primary system.
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3.
Offgas grab samples l
4.
N-16 activity measurements using portable steam line monitors !
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5.
Primary Leak Rate Calculation i
These methods are summarized in Table X-1.
j RMA-5 will be the first indication of increased primary to i
secondary leakage. The monitor will continuously sample the e
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.s
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vacu'um pu=p exhaust from the main condenser..Upon a 25: in-crease in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in RMA-5 count rate the primary and secondary syste=s will be sampled for tritium and the leak rate cal-culated. An of fgas'-grab sample will be taken and the primary to secondary leak rate will be calculated using Xe-133, Xe-135 and total gas activities. The portable steam line monitor will detect N-16 activity and will be.used to evaluate which steam generator is leaking.
Primary leak rate calculations which are done daily per Technical Specification requirement can also identify increased primary to secondary leakage.
Since the leak rate cannot dis-i tinguish,between unidentified system leakage and primary to secondary leakage, if an unidentified increase in leak rate occurs, a tritium and offgas grab sample will be taken to allow l for a'n accurate determination of the primary to secondar.y leak rate.
Shutdown limics' based on primary to secondary leakage will con-sist of the Technical Specification limit of I gpm and an admin '
istrative limit of 6 gph above a baseline leakage. Baseline leakage will be determined during the precritical hot testing program. When a leakage increase of 6 gph is reached the plant will be brought to a cold shutdown, the OTSC will be leak tested,
and the leaking tubes repaired. Tube leakage will'be tested by {
s the bubble test method.
This method has a sensitivity of
.1 gal / day / tube or 4x10-4 gph/ tube, therefore if no leakage is j
de'tected during the bubble test it can be assumed that no individual tube has reached the critical crack size and primary !
to secondary leakage is due to aggregate tube leakage. The baseline leak rate value will be redetermined based on an evaluation of the OTSC leak rate test results and operating history after the leak test is performed. When primary leakage i
reaches 6 gph greater than the new established baseline the j
plant will again be shutdown and leak tested.
j When shutdown is required by steam generator tube leakage, the plant should be shutdown expeditiously but in a manner to pre-clude reactor trip and subsequent lif ting of relief valves or atmospheric dump valves.* Cooldown rates should be limited to j
100*F/hr and tube to shell delta T should be limited to further ;
I reduce the possibility of tube rupture during cooldown.
1 Radiological Concerns B.
During normal cperation with steam generator tube leakage, radiological concerns arise in the following areas:
1.
General and specific area radiation level I
I 2.
Turbine building sump activity (with respect to discharge to -
,8 the environment).
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3.
Powdex resin a'n3 backwash water activity.
i Specific and General Area radiation li=its vill be deter =ined and vill be based en preventing the turbine building from beco=ing an RWP' area'-igreater than 5 cr/hr).
Limits are needed j
due to the necessity for easy access into the turbine building i
during operation.
Routine radiation surveys vill be taken in the turbine building in the vicinity of the Powdex and Graver i
System vessel and in other selected areas.
These areas vill be l
restricted if necessary to prevent unnecessary exposure to plant personnel.
Precautions vill also address secondary side system vent and drain operations.
In the Powdex su=p, pH and conducti-city analysis vill deter =ine if the water which has been processed by the (Ecodyne Graver) l Powdex Backwash Recovery system vill be returned to the THI-1 condensate system or to the turbine building sump.
Any radio-active powdex vill be devatered in Eigh Integrity Con-tainers / liners and shipped td commercial lov level vaste burial sites.
C.
Secondary Side Che=istry Secondary side chemistry li=itations for Boron and Lithium vill be based on considerations of chemical introduction into the.
turbine.
D.
Develorment of Procedural Guidelines for Steam Generator Tube Rupture A program has been formulated for providing i=preved operator guidelines for dealing with tube leakage and tube rupture'The The guidelines cover two categories of events.
events.
first category addresses tube ruptures for which subcooling margin is maintained. The second category will deal with tube ruptures for which subcooling margin is not maintained and would include various. contingencies including cultiple tube ruptures in one or both SG's, loss of reactor coolant pumps and loss of condenser.
1.
Contingencies for Cc6 sideration The following is an cuttine of the programs '*fo'r developi6g guidelines for SG tube rupture.
Ouidelines for Tube Rustures for Which Subcooline Margin a.
is Maintained The program to develop guidelines for tube ruptures for which sdbcooling margin is maintained vill include the following basic assumptions.
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TABLE X-1 LEAKAGE DETECTION METHODS SLW.ARY TABLE Method Sensitivity Frecuency Soecial Actions RMA-5 0.48 sph with 3.8 uCi/cc continuous strip W en count rate and 20'cfm exhaust flev chart reading.
increased by 25%
in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, sample for tritium and take off gas grab
- sample, j
t
.t
.3 gpm at.02 uCi ml 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with
/
Tritium H3 after 8; hours known leakage Offgas Grab
.01 gpm at 3.8 uCi/cc 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequ'ency increased Sample and 20 cfm exhaust flow with known leakage j
' Portable When leakage is 8
i Steam Line detected deter-1 Monitor mine which generator is leaking l
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'(1) Break si:e small enough to maintain sub' cooling margin.
(2) One OTSG affected.
(3). Reactor'-Coolant pumps operating.
(4) Condenser available.
(5) Decay heat removal from the non-af fected S0.
(6) SG steamed at 95% operating range level to assure natural circulation.
Contingency consideraticas for design basis tube ruptures' include:
(1) PORV unavailable.
(2) Reactor Coolant Pumps unavailable.
(3) No condenser available.
(4) High radiation release considerations.
(5) Steam line flooding consideration.
(6) Both SG's are affected.
b.
Guidelines for Tube Ru$tures For Which Subcooling Margin is Not Maintained The program to develop guidelines for tube ruptures for which subcooling margin is not maintained will include the following basic assumptions.
(1) Break size from one 'SG large enough to cause' loss of
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subcooling.
(2) No reactor coolant pumps running (since subcooling margin is lost).
(3) Condenser available.
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(4) PORV available.
(5) Unaff.ected SG is ste'med.
a Contingency considerations include:
(1) PORV unavailable.
(2) RCS voiding keeps pressure above SG safety valve setpoint.
(3) Primary. feed and bleed heat removal.
(a) With PORV available (b) Without PORV available Both the analyses employ the RETRAN code. This code models TFE-1 and has been benchmarked from transients on both IMI-1 and TMI-2.
Use of this code ensures that plant response under various primary-to-secondary leak scenarios is under-stood.
FIGURE X 1
-l Tube Rupture Guidelines Primary to Secondary f.eakage
> SO gpm J
A 4
l Manual Automatic Shutdown Shutdown t
1 Y
W l
Cooldown 3
v v
Forced Natural HPI l
Circulation Circulation Cooling 1
L J
l v4 Decay Heat Removal New Guidance:
- multiple tube ruptures
- ruptures in both steam generators
- HPI cooling
- Secondary water management l
r.
Improved guidance
- Minimum subcooling reduced
- RCP trip criteria
- tube to shell AT
- steam generator steaming, feeding, flooding 9
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The guidelines ' developed fron 'the RETRAN analysi.s for tube ruptures are su==arized belew and have been used for writing new procedures and revising old procedures.
Operator training prior t,o restart includes response to tube rupture events using new and revised procedures.
2.
Guideline Su==ary The sy=ptoms of a tube rupture define entry conditions for this emergency procedure.
It is only used when leakage exceeds 50 spm. When conditions require it (as defined by high leakage or significant of fsite releases), the plant will be shutdown and cooled as expeditiously as possible, and certain nor=al plant li=its (hCP NPSH, normal tube /shell delta T, and fuel-in-compression limits) are waived.
a.
I= mediate Action The tube leak in question may not*be large en~ough to cause a reactor trip. In such a case, the operator begins a load reduction as rapidly as possible without
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causing a reactor trip (10%/ min.).
Avoiding a reactor trip prevents lif ting of the OTSG safety valves, b.
Followup Actions (1) Subcooling Maintained and Reactor Coolant P.ucps Available Once the load reduction is initiated or 'a feactor trip has occurred, the operator has several major goals to achieve while bringing the plant to a cold shutdown condition. First, he prevents lif ting of the OTSG safety valves; second, isolates the af-fected OISC to prevent unnecessary radioactivity releases; third, minimizes primary to secondary leakage by minimizing primary to secondary dif-ferential pressure;.and, fourth minimize stresses on the OTSG tubes by limiting tube /shell delta T.
Finally, the operator will minimize offsite dose by allowing the *1eakage 0TSG to flood if of fsite, doses
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i are large enough (approaching levels, at which q Site Emergency would be declared).
The major differences between the existing plant procedure and the new procedure would be the following.
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,I (a) Maintain a Minimum Subcooling Marrin Mini =u= subcooling margin =eans that pri=ary to see.ondary dif ferential pressure is cini=i:ed.
Mini =um differential pressure means that leakage is reduced ; thus reducing off site dose and =aking the event more manageable.
In order to maintain the =inimum subcooling margin, several plant limits have to be violated:
fuel in compression limits and RCP NPSH limits.
The former is acceptable to violate during e=er-gency conditions, while the latter is being reevaluated to determine acceptable emergency operation of the. pump.
j t
(b) Steaming / Isolation Criteria for the Affected OTS G The present procedure allows the ope'rator to j
let the OTSG flood anytime that RCS pressure is I
below 1000 psig.
The revised procedure has the operator steam the OTSG as necessary for the following purposes. First, to prevent lif ting of the OTSG safety valves.
As the OTSG 1evel increases, steam generator pressure in the is'olated generator could incre::e toward the r
safety valve setpoints.
Pressure should be controlled to prevent a safety, valve lift.
The generator is also steamed to preve'nt it from flooding.
Flooding is undesirable because an RCS pressure increase under this condition could cause water relief out of the OISG safety valves' A flooded OTSG would also act as a second pressuricer and slow depre::urization of the RCS (as occurred in the GINNA tube rupture).
I I
l The OISG will be isolated under two conditions.
First, if BWST level goes below 21 ft. indi-cated level.
At this level, there is still sufficient inventory to fill up both OTSQ's to the main steam isolation valves aod have QO,00Q gallons of water left to go on feed and bleed i
8 cooling.
A second rea. son to isolate the OISG is for radiological considerations.
If the Radioiogical Assessment Coordinator (RAC) l determines that of fsite doses are approaching j
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th'e levels which'uculd require declaration of a Site E=ergency. (regardless of cause) the affected stea: generator will be isolated.
(c)
Tube-to-Shell Delta T Plant ad=inistrative li=its and precautions will require caintaining the OTSG average tube temperature within 70'F of the average shell te=perature.
Under emergency conditions, this li=it can be relaxed to 140*F (Tech Spec li=it) withou,t adversely affecting O!SG tubes.
,i (2) Loss of Subcooling Margin with Natural Circulation Cooling l
When RCS subcooling is lost, the operator,cust treat LOCA, as well as tube rupture symptons.
First he trips RCP's and then verifies' EPI and EFW have initiated. He is then able to pursue the followup tube rupture actions. All of the guidance for followup actions without loss of subcooling apply, as well as the additional guidance provided below.
L The objective in this portion of the procedure is to maintain natural circulation, reestablish subcooling l
=argin, restart a reactor coolant pu=p, and return l
to the section of the procedure for foteed flow cooldown.
When subcooling is regained in the RCS, then HPI is throttled, RCP's are started and the operator, con-tinues with 1.67F/cin cooldown.
If subcooling can-not be restored, the operator enn1= the plant devn on natural circulation, stea=ing as necessary to meet the objectives described in the forced flow section.
If the affected OTSG cannot be steamed for either radiological or equipment reasons, then EEW is used i
to control OTSG pressure.
Essentially, EFW is used as a pressurizer spray to keep the leaking ge'nerator slightly lower in pressure than the "ECS.
The behe
- fits in controlling steam pressure are:
(a ) safeties will' not lif t.
(b ) the steam generator will not control RCS pres-
,sure.
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l (c) there will not be backleakage into the RCS of a
unborated water.
(d) leakage from the RCS to the OTSG will be small i
since dif ferential pressure will be small and will also reduce tube tensile load due to pres-sure loads.
(e) the small flow through the hot leg will help prevent void formation in the hot leg.
(3) Loss of Subcooling and Loss of Heat Sink Natural circulation cooldown will continue until subcooling is restored or the OTSG heat sink is lost
'(for example, due to loss of natural circuldtion in r
the unaffected loop). With no steam generator heat sink, the operator must put the plant in' a feed and l
bleed cooling mode. Feed and bleed cooling is ini-tiated by isolating the OTSG's, assuring full HPI is j
operating, and opening the PORV.
If RCS pressure l
remains below 1000 psig, then the operator continues a
to control. secondary side pressure just below RCS pressure.
If the OTSG heat sink is restored, the l
feed and bleed is terminated and a natural cir-culation cooldown is reinitiated.
l If RCS pressure stays above 1000 psig during feed and bleed cooling (e.g., the head b'ubble p'revents depressurization or the PORV fails closed) then the secondary side safety valves have to be protected from challenge. The operator controls OTSG pressure with whatever means are available (turbine bypass, EFW or ADV). When the OTSG is about to flood, the operator opens the ADV and leaves it open. This action minimizes the chances that safety valves will i
be forced ec relieve water and/or steam and fail j
open. The steaming capacity of an ADV at 1000 psig i
exceeds decay heat levels within several minutes i
af ter reacto'r trip. HPI capacity exceeds the capacity of one ADV.
Therefore, the RCS press ~
can be controlled at 1000 psig in this' mode wi,ur,e j
thout lifting safety valves.
Subcooling margin can be regained and the p,lant cooled down in this mode until an OTSG heat sink can be restored or until the plant can be put on decay heat removal.
A simplified schematic of the tube rupture guide-lines is shown in Figure X-1.
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A fou,rth'possible scenario exists under current procedures which has not been considered a preferred course of action in for=ulating the guidelines:
=aintenance of subcooling margin but tripping of -
reactor coolant pu=ps on 1600 psi RCS pressure.
Pu=p trip on loss of subcooling =argin instead of RCS pressure allows the operator to maintain forced ficv for about 3 ruptured tubes - 1600 psig S?AS is much more restrictive.
Forced RC flow provides several benefits during a tube rupture.
1.
It minicizes primary to secondary delta P and thus reduces tube leakage and tube tensile load.
2.
Prevents steam formation in the RCS.
(Steam voiding prevents RCS depressurization.).
3.
Provides pressurizer spray so that RCS pressure control is not dependent on the PORV or pres-surizer vents.
Therefore, CPUNC is taking action to have the 1600 psi pressure pump trip requirement changed to trip on subcooling cargin.
E.
Conclusions Primary to secondary leakage vill be monitored during non-steaming and steaming conditions.
Sacpling require =ents on the detection of a primary to secondary leak have been established, and ad=inistrative limits on leakage are being considered.
The combination of analysis of tube ruptures, procedure improve-cent and training ' improvement give assurance that operators can safely respond to a primary to secondary leak.
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