ML20092K781
| ML20092K781 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/25/1984 |
| From: | Churchill B METROPOLITAN EDISON CO., SHAW, PITTMAN, POTTS & TROWBRIDGE |
| To: | Hetrick D, John Lamb, Wolfe S Atomic Safety and Licensing Board Panel |
| References | |
| 83-491-04-OLA, 83-491-4-OLA, OLA, NUDOCS 8406290102 | |
| Download: ML20092K781 (16) | |
Text
@
7_,__
-rm:DEC SHAw, PITTMAN, PoTTs & TROWBRIDGE a pamTNEmSMep CF poorESSiONAL Compomat ONS 1800 u simEtt. N. w.
WASHINGTON. D. C. 20036 s
Am C 26Ma m u e nowaN.
nama-wo m.aMSa, DnTM.D e
,0 N E.G E 6.C.e.... p C DE.e ma.m.
S.
p
.A S Ema n? < p
~. p C C.TMOM SM S
SCO A Eh.EmG
.E. NET...mte GE-mGE r vmOwenaDGE. p.C.
C M amL E S S. TE **m eh.
- C,
~
CAMPBELL msLLErge wENDELew a woesTE L
T pN e we
- C TE LECO"'E N p"
3"m "
i E g. "S ee CM F,p g
M M
se M
ti?"o 1 Mit"s2"@':"" *.
. rJ"Tt'"a',?E ""C,4lfE
~~hr TS.*fa*J"dc">"C.
n"1* *."*"i'."4 *-
aOm emose a em'ee in"":@"'?a' ny"C' lt; 3",Pa*Oh1
? P"1E "TT"i'E"c' ""'c "8:l0 f.*O".'!"cS"RC.
L w.N
[Q L S'amDea E sOa*N s.mMehELAh0Cn.PC.
STEVEN M LvCA S.
Dawso ** muet= STEIN.*C maatFA A ICO mENM&TM J MAWTMAN S
.sawSCA b
T'esek. m at a vec la J mE A NS.
(2021 822:072 STEvt =,CR L m DE MLEiN p e La M A DERS N meCesanD J mE hDALL P C.
JOMN M O mEi6 6. s m.. p C.
psTLE n aLEmamDEm D TC'heaSICZum
%a YmEDEL (kun Cbw I a 6SSOTTI. D C.
JNFEnvL. A S'ko, p C.
E L
- E N a
GEOmOE y ALLEN,sm..*C.
J ACM macnav
- C TELEz mosERT C e O E M **
- M'Les D. pomTE m i mE D DeaS=E n. P C atCManD E. GALEN Mam+E h.COLahD McCMAE L a. SweGER ha E p enEED m.. P C RT **
LASS
- 6 D ANE Sun LET Ta J N E STEIN an ' an avoENGLecm.
- C TMOMASM M -Co mad eCm CASLE *SMawbaw' L. DwahE CME E m EtLEEN M G L EiM g m Na\\f*O'N ONYS **C CiIan VMOmas a sAsTER,a C-staN me. G A LLO* AT s"y 0 M"a OwDO N"TealN S
> NTANGE S N Em EDwa#D D. vow =G. til sEFrmEv w.maMpCLMan sam.S M Sum 0E n. p C.
Jose ** 6. Ca m m. J m.
JOMN r. DEALT
- ahontw D ELLIS RE**NETM a. G ALLO J ?
M EN A T.
Am MOR EN V CM m Mand G Dev D n av N SO**mst S COTTL E B COwhSEL ptTE R J FEmmaAa m* Casa #D K a SECpta t.TEE N L. M E LT2 E R. p C.
L En N D. AWLtC A, P C.
MowanD M. SMAF'F E RM AN MELEN TORELLI wlLSERT WASMihGTON se June 25, 1984
.m.. E
- w.,E n S D.mECT D,ak hwM.E m (202) 822-1051 Sheldon J. Wolfe Dr. David L. Hetrick Administrative Judge Administrative Judge Chairman, Atomic Safety and Atomic Safety and Licensing Board Licensing Board Professor of Nuclear Engineering U.S. Nuclear Regulatory Commission University of Arizona Washington, D.C.
20555 Tucson, Arizona 85271 Dr. James C. Lamb, III Administrative Judge Atomic Safety and Licensing Board 313 Woodhaven Road Chapel Hill, North Carolina 27514 In the Matter of Metropolitan Edison Company, Et Al.
(Three Mile Island Nuclear Station, Unit No. 1)
Docket No. 50-289-OLA and ASLBP 83-491-04-OLA (Steam Generator Repair)
Dear Administrative Judges:
For the information of the Licensing Board and the parties, I am enclosing a report prepared by the Subcommittee on Steam Generator Repair of the TMI-l General Office Review Board (GORB).
The report deals with the repair of the steam generator tubes.
Sincerely,,
'[
8406290102 840625 y
PDR ADOCK 05000289 O
PDR Bruce W. Churchill BWC:smm Enclosure
~0 cc:
Service List t
~C C
'OA E'25 m:54 TMI-1 GORB SUBCOMMITTEE ON STEAM GENERATOR REPAIR REPORT TO THE TMI-1 GENERAL OETICE REVIEW BOARD 13 June, 1984
/
S V
i T."L. Gerber, e
A D. T. Lei httfli, Member Y4E Y/55 14 W. Lowe, Subcommittee Chairman
I s
s j j TABLE OF CONTENTS i SEL*fION PAGE I.'
BACXGROUND 1
5 g..
II. /' CONCLUSIONS 3
III. (AUSE OF bTEAM GENERATOR DAMAGE 4
s IV.
' INSPECTION FOR STEAM GENERATOR DEFECTS, REPAIR
.j,0f ' DEFECTS, 'AND INSPECTION OF THE REPAIRS 5
V.
DISTRIBUTION OF SULFUR S
VI.
CHEMICAL CLEANING TO REMOVE SULFUR 6
VII.
EVALUATION OF THE INTEGRITY OF REPAIRED STEAM GENERATOR TUBES AND DEIECTION OF LEAKS 6
VIII. ANALYSIS OF PLANT RESPONSE TO STEAM GENERATOR TUBE LEAKS / RUPTURES, AND ADEQUACY OF OPERATING AND EMERGENCY PROCEDURES AND TRAINING OF OPERATORS TO CONTROL TUBE LEAKS / RUPTURES 8
IX.
CONTROL OF FLUID CHEMISTRY AND PLANT CHEMICALS 12 e
9 e
y,
r-
I. BACKGROUND tests in August and SeptembeIn late November 1 ng hot functional _
tubes in both, steam r, a large number of found to be leaking. generators of TMI, Unit 1 w the 15,531 tubes in each gemeasurements ind ere eral thousand of circumferentially from the insidenerator were cracked cracks were near the top of the t b Most of the tubes are in the 24" thick u u es where the Fall of 1982 extensive additipper tube sheet.
discovered at the top end of thona* cracking wasIn the extend a fraction of an inch ab e tubes where they Destructive examination sho ove the tube sheet.
j intergranular.
Sulfur compounds in significantwed tube crac the inside of tubes and othamounts wer ure surfaces and on
- surfaces, er primary system interior 40% thru-wall indications belUltimately about 120 I
greater than All tubes not plugged were eunderside of the uppe were plugged.
for the seal requires that it bseal them against th The specification in length and below all known d fe six inchas or as found b more expansion.y eddy current tests prior to explosive e ect indications done which show that such aTest data were taken tightness and structural strseal provides leak yses were original joint.
ength equivalent to the initiatedFollowing the discovery of c the cause,a comprehensive investigation to establi racking, GPUN to the steam generators andto measure the extent and nat t
sh investigate methods for other equipment, toure of damage and operation of the unit. conditions for the repair, re.ir repa certification, restart on all of these subjects.
GPUN has issued reports TMI-1 General Office Review BThe Subcom 1
erator Repair of the on 22 April 1982. appointed by the GORB Chairman at GOR oard (GORB) was draft an opinion for the GORB'The Subcommittee was asked to3 Meetin TMI.-Unit 1 as it might be affconcerning the safety c art and operation of the steam condition. generators and by their as-repairedected by the repair of 7013G061384 1
4 April 1982 for a total of more thThe Subcomm times sin
.has had more than 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> of presentatioan 200 ho Nuclear Personnel on matters rel It n by GPU pertinent documents prepared by GPUNass ating to its access to contractors.
and its written requests for informatiFrom time to time th ee has made and commented on the developing stmeeting on.
, it has questioned about the steam generators and upo ate of knowledge or taken by GPUN, n actions proposed Subcommittee proposed a GOPJ3 rat TMI-1 GOR ugust 1982 the concerning control of plant chemist ecommendation recommendation, as subsequently re i ry.
This by the GORB in telephone meeti v sed, was approved GORB Recommendation No. TMI-1-50 1ng 20A and issued as background statement to be a fundconsiders th n
the associated representation of its views and in amental reference as part of this reportcorporates them by an official response to the recom.
The GCRB received December 1983 meeting.
mendation at its 14 On 8 September 1982 the Subco generator repaconcurred with the GPUN finding th tmmittee form repair process,ir gave sufficient assurance that theplans fo a
1983 the Subcommittee issued ain itself, was safe.
On 10 August GORB concluding that subject to thn interim report to the operation would not be adversely affseveral m restart and repaired steam generators.
ected by the Comnittee forwarded an opinion ton 22 August 1983 the that the full pressure hot fun tio the GORB Chairman generators could be conducted withonal test of the c
the public.
minimal risk to The Subcommittee's conclusion c safety of operation with repaired st oncerning the stated in Section II, along with th eam generators is which it is based.
e provisions upon numbered items at the end of SectiSubcommittee IN.
pear as indented viewed by the Subcommittee as beiConsideration o ations is not for power operation.
ng a prerequisite 7013G061384 2
II. CONCLUSION the Subcommittee concludes that TMI, Unit I can be safely operated with the repaired steam generators provided:
(1) on-site capability is established and implemented for measuring with a short turn-around time reduced sulfur compounds in primary water at and below concentration limits considered to be low enough to prevent damage (Section IX).
Such capability should be implemented as soon as feasible, but in any event before the primary system is i
exposed to air after power operation:
(2) prior to_ power operation, operators (a) are trained to understand all plant instruments and symptoms which could be used to indicate subcooling margin and loss of subcooling, (b) are required to determine subcooling margin using at least two different instrument sets I
when controlling plant depressurization and cooldown following a tube break and (c) are trained to understand what would happen in the plant and what they would observe if the error in a subcooling margin measurement is larger than actual subcooling margin in the plant and they were using such a measurement to control subcooling (Section VIII (2));
(3) Prior to power operation, GPUN complete and evaluate an additional steam generator i
leakrate test to be performed by plant j
operators, as presently planned, utilizing injection of krypton into the primary system.
i f
(4) results from additional hot functional tests, start-up tests and steam generator inspections planned 90 to 120 days after l-power operation do not show signs of significant tube degradation (Section VII).
l l
i l
I 3
7013G061384 I
~
III. CAUSE OF STEAM GENERATOR DAMAGE i
A program to investigate the cause of extensive cracking in the Inconel 600 tubes of the TMI, Unit 1 I
steam generators was undertaken by GPUN and their,
contractors soon after discovery of primary to secondary leaks in November 1981. Major elements of this program included (1) review of the fabrication history;-(2) in situ and laboratory examination of
(
tubes and tube cracks; (3) review of plant water j
chemistry and operating history; (4) tube stress e
analysis, and (5) development of a failure scenarzo.
In addition, an extensive corrosion test program was j
undertaken to investigate the cause of cracking and the performance of repaired tubes with and without chemical cleaning.
The Subcommittee is satisfied that the cause of cracking has been thoroughly investigated, and that the most likely failure scenario has been defined.
This scenario indicates the cause of tube cracking was the presence of reduced sulfur compounds in the l
i primary water which created an environment conducive to intergranular stress corrosion cracking of tube material at the static air-water interface in the
-upper part of the steam generators when they were partially drained during shutdown following hot functional testing in August and September 1981. The tubes were sensitized during heat treatment of the steam generato:s. The majority of cracks occurred j
where the tube material tended to have locked stresses in the upper seal weld region and the roll transition zone.
During the course of the Subcommittee's review, a number of questions related to the cause of cracxing were posed. The majority of these questions were concerned with understanding the cause of cracking in sufficient detail so that potentially related problems could be identified, and the safety of future operation could be assessed.
'A related objective in some of the Subcommittee's questions was to understand how to reduce the chance of future l
damage due to inadequate chemistry control.
[
There is. strong circumstantial evidence to support the failure scenario proposed by GPUN.
Laboratory simulation of a range of conditions shows
~
stress corrosion cracking of the sensitized Inconel t
600 tube material will occur at relatively low temperatures when it is highly stressed and is i
4 7013G061384 n.-
,,,ny--ny-
-,,,--n
-e
-.-,,,,,,n,
,...,,. + -, - - -,~ ~
-e,-
--~,, -.-
submerged in aerated water containing reduced sulfur compounds (e.g., thiosulfate ion) at concentrations as low as one ppm. There is a high probability sodium thiosulfate was inadvertently introduced into primary water from the borated water storage tank the contents of which had been contaminated with thiosulfate bearing water from the containment building spra;r system.
IV. INSPECTION FOR STEAM GENERATOR DEFECTS, REPAIR OF DEFECTS, AND INSPECTION OF REPAIRS As they were being developed and implemented, the Subcommittee reviewed the programs to identify and characterize steam generator tube defects, to repair the tubes and to inspect the repairs.
Based on its review, the Subcommittee concludes that the inspection, repair, and reinspection techniques developed and implemented by GPUN and its contractors and subcontractors have been thiscough and have properly utilized state of the art techniques available.
V. DISTRIBUTION OF SULFUR Circumstantici evidence indicates sulfur-induced intergranular stress corrosion cracking damaged steam generator tubes, pilot operated relief valves, gaseous waste piping, and spent fuel pool cooling pipe. Sulfur compounds were found in the cracks as well as inside the pressurizer and at other locations in the primary system. From all this it is inferred l.
that sulfur compounds were present throughout the primary system and in some connected systems.
The spent fuel pool cooling pipe was repaired several years agb, and no further deterioration has been detected. The TMI-l pilot operated relief valve I
(PORV) was replaced with a spare valve.
The corroded PORV was cleaned, repaired, and subsequently i
reinstalled.
In early 1984, the valve was removed, inspected, and found to be in good condition.
The l
gaseous waste piping has been repaired.
The steam generator tubes have been repaired or plugged as required.
l 5
l.
7013G061384 l
i
~
An extensive examination by GPUN was made to determine whether damage other than that identified above had occurred in the primary system or in systems connected to it.
These examinations have not y
revealed any additional damage celated to sulfur
~
induced corrosion.
i VI. CHEMICAL CLEANING TO REMOVE SULFUR i
A hydrogen peroxide chemical cleaning process has been carried out P.o rePrve sulfur from the primary system.
Prior te chem:: cal cleaning, internal surfaces of the pressurizer were cleaned with high pressure water jets.
The chemical clear;ng process was designed to convert residual radu<.1d sulfur species in surf ace deposits to soluble u 1 fates, and to remove trem from l
primary water by ion 3xchange. Tests on TMI-1 steam' j
generator tube camplas by Battelle Columbus Laboratories irdicat:0 that reductions of 50 to 80 percent of.the sulfue species could bs achieved. No harmful effects were identified by the tests of the cleaning process.
While the Sube;mmittee recognized that there were uncertainties as to need and effectiveness, it believed it praden'c to p'roceed with the cleaning.
It reviewed the plans and specifications for cleaning and found them adequate.
i-VII. EVALUATION OF THE INTEGRITY OF REPAIRED STEAM l
GENERATORS lI i
GPUN has :arried out an extensive program to establish the integrity of the repaired steam generators and to demonstrate that they can be l
operated safely. The program has four essential ob,iectives.
The first is to qualify the repaired areas to original design criteria. These areas include the explosively expanded tube to tube sheet l~
joint, the plugged tubes and the upper ends of the tubes which were repaired by machining and cleaning.
'The second objective is to verify that unrepaired sections of steam generator tubes below the expanded joint in the upper tube sheet do not now contain cracks which could rupture in normal operation or design basis transients or accidents. The third
\\
6 7013G061384
~
mechanism which caused the damage h e corrosion and will not reactivate.
demonstrate that, if tubes deteriThe fourth is toas been future, leakage through them will binterg orate from c ing in the sufficient time to shut the plant doe detected in breaks.
wn before a tube The first two objectives of thi been achieved utilizing repair mocks program have steam generators, laboratory testsanalyses
-ups, detailed ortions of the comprehensive hot functional testsgenerator r repair and corrosion mechanism has been arrestedThe 1
g that the reactivate, has been addressed in se and will n Laboratory tests and chemical thveral ways. ot of sulfur compounds have been made termodynamic analysi specifications for primary coolant ch o establish conditions known to cause damage will bemistry so that under operating and shutdown modes e avoided tests of the repaired and cleanedHot functional have been and will be conducted t steam generators deterioration by measuring prima o detect tube rate.
out which simulate plant operating cLong-term ng carried are experienced in the plant. Test results to dateycles before the it will occur under conditions spshow no evi n
nued or that operation.
ecified for future To serve the fourth objective established a primary to secondary leak
, GPUN has administrative limit of a nominal 6 g ll rate (GPH) above a baseline valu a
ons per hour during power operation as determie (currently one GPH) i radioactivity in the ned by noble gas in condenser off gas. primary coolant compared to that If the leak rate exceeds the nominal value during steady state op plant will be shut down and the steam eration, the inspected.
i measurements made during the cooldowAfter each pla generators GPUN is required by Technical Spe ifievaluate ae sumed. Also the plant to cold shutdown if the c
cations to bring unidentified leak rate exceeds 1 galloprimary system (60 GPH).
n per minute.
L i
7013G061384 7
The Subcommittee believes that proposed leak rate limits and plans for evaluating cooldown data are appropriate and have a high probability of identifying significantly degraded tubes so that the unit will be shut down before a tube break due to intergranular stress corrosien cracking could occur.
Based on its review of the information and documentation in support of the four objectives, the Subcomittee recommends that:
(1) GPUN document the updated description given orally to the Subcomcittee about the logic, analyses, and leak rate measurements which provide the bases for concluding that leaks from significantly degraded tubes will most likely be detected before one of them breaks. Clarification is needed because statements in early GPUN reports and quantitative analyses, when viewed individually, imply conclusions which are more definitive than can be supported by current state of the art analyses.
Such clarification is needed so actions taken to assure safe operatien (including primary to secondary leak rate ceasurements) can be kept in proper perspective.
VIII. ANALYSIS OF P'.R.T RESPONSE TO STEAM GENERATOR TUBE LEAKS /R"PTURES A*ID ADEQUACY OF OPERATING AND EMERGINCY PROCEDURES AND TRAINING TO CONTROL TUBE LEAKS / RUPTURES.
The Subcommittee has recognized from the outset, as has the GPUN staff, that tube ruptures or large leaks may occur due to future degradations or to existing conditions not now recognized.
Therefore, the acceptability of future operation of these or any l
other nuclear steam generators must rely on confidence that operators will be able to shut the plant down before leakage creates a hazard to people or damage to the plant. This confidence has to be founded on a judgment of how tubes can break, on analyses of how the plant responds to tube breaks, on an ability to detect leaks which warn of incipient I
breaks and, most importantly, on the adequacy of operating and abnormal transient procedures and l
operator skill and training. Consequently the i
Subcommittee has reviewed these matters in detail.
8 7013G061384
Although GPUN is continuing to refine their analyses of plant response to abnormal transients such as postulated main steam line breaks and the related effects on steam generator tubes, the Subcommittee is satisfied that sufficient data has been developed through analvses and simulator drilla to validate the approach used to control tube leaks (less than 50 gpm) and tube ruptures (50 gpm or greater) as specified in the new TMI-1 Abnormal Transient Procedures (ATPs). Based on its review and provided the training cited in paragraph (2) of Section II is completed, the Subcommittee considers that the present state of the procedures and operator training is sufficient to provide a high degree of assurance that the operators can safely handle tube leaks and ruptures should they occur, including leak rates from multiple tube ruptures which exceed the design basis by a significant amount.
The Subecmmittee recommends that the following points be considered for procedure revisions,
_ operator training, plant testing and/or analyses:
(1) An independent verification should be made of the error analysis which supports selecting l
the subcooled margin limit applicable to reactor coolant pump. trip following a steam generator tube break. The verification should be equivalent to that required by ANSI Standards for design verification. The analysis and the verification should be subject to an interdisciplinary technical review by senior engineers who understand error analysis and primary system response to tube breaks and other loss of coolant i
accidents, l
L (2) For tube rupture transients,the Subcommittee I
concurs with the desirability of reducing indicated subcooling to 25'F (provided this value is verified as noted above) or to emergency RCP NPSH limits (whichever is more limiting).
However, the possibility exists that an instrument string which measures l
subcooling margin could read erroneously high by more than 25'F.
If.that were to happen,it is important that the operator recognize that the instrument is in error, since once actual l
saturation has been reached the instrument l
reading will " hang up" at the value of error l
(higher than 25'F in this postulated case)
I' while the operator continues to reduce l
l I
9
(
7013G061384 L
i pressure. The instrument will not indicate lower saturation margin until superheating from uncovered fuel commences. Since there are several independent instrument strings which can be used to measure subcooling, the Subcommittee does not consider it credible-that they would all read erroneously high by more than 25*F at the same time, provided calibration procedures and equipment preclude a common error in all the instrument strings. Therefore, the operator should be l
able to avoid reaching saturation conditions without knowing it. However, it is important that all operators understand the symptom (described above) which would be observed in this case and that they understand which other plant symptoms would indicate saturation had occurred (i.e., changes in pressurizer level, pump current, etc. ). The i
Subcommittee recommends these matters be addressed in the training discussed in Section II (2).
(3) In the event of a tube rupture, operating the reactor coolant pumps as long as feasible (without incurring real troubl-if they are subsequently shut off) can reduce total
-leakage through the break. Therefore, if the results of ongoing analyses show that the time limits for reactor coolant pump trip can be extended, the current limits should be reconsidered.
(4) The new fuel pin in compression limits released by B&W should be incorporated into TMI-l operating procedures as soon as practicable. These limits will apply to normal heatup and cooldown and tube leak j
(less than 50 gpm) transients. Comparison I
of these limits with the cooldown data in l
TDR 488*shows that the new fuel pin in compression limits will allow much greater l
depressurization of the RCS before a cooldown l
results in high tube tensile stress due to tube to shell differential temperature.
Since any tube cracks can be expected to open wider as the tube to shell differential temperature increases during cooldown, it is important that the tube leak procedure (ATP 1210-5) and/or the training program stress that in responding to a tube leak (less than 50 gpm) priority be given to minimizing 10 7013G061384
primary pressure within allowable limits before the cooldown results in high tube to shell differential temperature. This will reduce the total leakage during the i
cooldown. Further, if the leak then develops into a rupture (greater than 50 gpm) as the-tube tensile stress builds up (as happened at Rancho Seco.in May 1981) the resulting leak rate will be less, and the time needed to reduce the RCS pressure to RCP emergency NPSH limits allowed under tube rupture conditions will be less. Also, the RCS pressure is more li)>ely to be below the secondary side safety va've lift pressure before the rupture occurs.
(5) Use of the new fuel pin in compression limits would permit plant testing over a wider range of temperature to verify the RCP emergency NPSH limits applicable to tube rupture and LOCA events.
Such tests should be considered.
(6) Since the procedure for a tube leak (less than 50 gpm) requires that fuel pin in compression limits and normal RCP NPSH limits apply, these requirements should appear next to each other in the procedure and both should refer to the same graph, probably Figure 1 and 1A in OP 1102-11 (Plant l
Cooldown). There would then be no need for fuel pin compression limits in ATP 1210-10.
i (7) The list of questions in the form for the Reactor Trip Report attac:ted to ATP 1210-1 should include:
"Were fuel pin in compression limits violated?" since, if they were, the data would normally be evaluated by B&W before restart.
(8) Guidance should be given in the ATPs and OP 1102-11 as to what actions are required if l
the cooldown rate exceeds 100*F/ hour for j
limited periods.
l (9) The Abnormal Transient Procedures should be clear and coherent to reduce chances of i
mistakes in handling tube breaks. These ten L
procedures could be improved if they were all i
made consistent in the following respects:
I a.
Consistently refer to ATP 1210-10 whenever a requirement from that procedure is required to be invoked to 11 r
i 7013G061384
i t
carry out a step in the other nine
. procedures.
At present some steps in i
some of the ATPs do this, but others do
- not, f
b.
Whenever "subcooled margin" is intended to mean "25'F subcooled margin", so state.
At present this is done sporadically throughout the ATPs.
i Use a consistent paragraph numbering c.
system.
i i
IX. CONTROL OF FLUID CHEMISTRY AND P the fuel pool cooling system, and the waste i
t system occurred because methods then in use for controlling water chemistry and the ingress of deleterious chemicals were inadequate i
how they could occur. understanding of potentially damaging
, as was the
}
i Since then GPUN has of plant chemistry. implemented extensive measures to im i
control sulfur compounds in primary water as describ I'
{
Section II, paragraph (1) is est blished, the Subcommittee considers current GPUN capability f a
. control of fluid chemistry and plant c4emicals, if i
or assiduously applied, is sufficient to keep the acceptably low level during start-up and powe i
operation.
However, the Subcommittee recommends that:
t i
(1) Concentration limits for reduced s lf
~
L be established for operating and shutdownco u ur l
t i
modes as soon as feasible at levels low
!~
enough to prevent damage.
This is not a prerequisite for power operation.
L (2)
That continuous on-line monitoring of important chemistry parameters in primary l~
L water be established.to the extent practical so that short-term variations in parameters can be detected, recorded, and, if they approach or exceed limits, alarmed.
t-This is not a prerequisite for power operation.
i 7013G061384 12 i
- +-
,,m..
.m
=
c.
~
~.,
UNITED STATES OF AMERICA
'84
~
NUCLEAR REGULATORY COMMISSION d'Y26 NO:54 BeforetheAtomicSafetyandLicensincfBoard
$1 In.the Matter of
)
)
METROPOLITAN EDISON COMPANY, ET AL. )
Docket No. 50-289-OLA
)
ASLBP 83-491-04-OLA (Three Mile Island Nuclear
)
(Steam Generator Repair) i Station, Unit No. 1)
)
SERVICE LIST Sheldon J.
Wolfe Atomic Safety and Licensing Administrative Judge Board Panel Chairman, Atomic Safety and U.S. Nuclear Regulatory CommissiC Licensing Board Washington, D.C.
20555 U.S. Nuclear Regulatory Commission.
Docketing and Service Section (3)-
Washington, D.C.
20555 Office of the Secretary U.S. Nuclear Regulatory CommissiC Dr. David L.
Hetrick Washington, D.C.
20555 Administrative Judge Atomic Safety and Licensing Board Joanne Doroshow, Esq.
Professor of Nuclear Engineering Louise Bradford University of Arizona Three Mile Island Alert, Inc.
Tucson, Arizona 85271 315 Peffer Street Harrisburg, Pennsylvania 17102
'Dr.
James C.
Lamb, III
-Administrative Judge Jane Lee Atomic Safety and Licensing Board 183 Valley Road 313 Woodhaven Road Etters, Pennsylvania 17319
. Chapel Hill, North Carolina 27514 Norman Aamodt Richard J.
Rawson, Esq.
R.
D.
5, Box 428 Mary.E. Wagner, Esq.
Coatesville, Pennsylvania 19320 Office of Executive Legal Director i
U.S.
Nuclear Regulatory Commission Washington, D.C.
20555
' Atomic Safety and Licensing Appeal Board Panel U.S. Nuclear Regulatory.Comission Washington, D.C.
20555 l
i'