ML20090B428
| ML20090B428 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/19/1982 |
| From: | Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Pollard B UNION OF CONCERNED SCIENTISTS |
| Shared Package | |
| ML082380335 | List:
|
| References | |
| FOIA-91-106 NUDOCS 8203310038 | |
| Download: ML20090B428 (18) | |
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b Od VJ y3 9 g Mr. Bob Pollard Union of Concerned Scientists Suite 1101 Dupont Circle Building 1346 Connecticut Avenue, N.W.
Washington, D.C.
20036
Dear Mr. Pollard:
I am writing in response to the questions asked in your letter of January 28, 1982 which addressed the recent stean generator tube leak at the R. E. Ginna Nuclear Power Plant and tha genertl problem of steam generator tube degrada-thn.
As a consequenc'e of the Ginna event, staff was reouested by the Commission to establish a Task Force to review and evaluate the event.
The report that I expect from the Task Force by April 2 will deal with some of your concerns
,in more detail than is presented in the attachments to this letter.
I trust this letter is, responsive to your inquiry.
Sincerely,
- (Sigr.ed) 2. Re vin Cornell William J. Dircks Executive Director for Operations
Enclosures:
1.
Chronology 2.
Responses to Questions 3.
Radioactive Material Releases cc w/ encls:
Richard Udell Paulette Meier Distributior.:
TDTlid74 DeYoung SECY 82-59 Cunninaham, ELD Dircks RHaynes. RI Cornell Ipcoli k j
Rehm Eisenhut Denton Revised in OEDO, see previous ORC for concurrences h
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Preliminary Event Chronolooy R. E. Ginna Steam Generator Tube Failure Time Event Comment 9:25 a.m., 1/25/82 Charging Pump sp'eed alarm; "B" Steam First indications Generator (5/G) steam flow - feed flow of tube rupture mismatch; air ejector radiation monitor in the "B" steam alarm; Pressurizer (PIR) low pressure generator, alarm (setpoint - 2170 psig).
Shift Supervisor orders power reduction:
Shift Supurvisor one operator fast closes turbine control (SS) and operator valves; another operator commences actions based on normal boration.
oral interview with the S1.
9:28 a.m.
Reactor trip on low pressure (setpoint -
Seal retun 1873 psig with a rate factor); automatic isolation valve safety injection with a containeent closed on'contain-isolation (setpoint 1723 psig); Reactor ment isolation Coolant Pump (RCP) seal injectier. return causing line line pressurfres eventually lifting its to pressurize.
relief valve: PZR levti rapidly falling; The contribution 5/G "A" and "B" lew level alarms, of this reitef to the PRT is believed insign-ficant.
The S/G low levels resulted from the combined effects of the power reduction and the reactor trip.
9:29 a.m.
Both RCP's manually tripped in accordance Station procedures with Station Emergency Procedures E-1.1 require tripping and E-1.4; RCS pressure about 1750 psig RCP's at 5 1715 I
and dropping.
psig; Westinghouse guidance specifies trip pressure as 1200 - 1300 psig for Ginna.
Licensee trips pumps at NOTE highar pressure due to pressure This chronology, developed by Region I on February 5,1982, instrument represents an improvement over that prepared on January 26, qualification l-1982 and will be used by the Task Force in the development status.
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Time Event comment 9:33 a.m.
NRC Operations Center notified by the Further discussion
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licensee via the ENS phone; the licenset revealed that the reported a reactor trip from 100% power licensee strongly as a result of a steam generator tube suspected that the rupture. The faulted S/G and release "B" S/G contained informatien was not given by the lice'nsee the fault, but at this time, the licensee chose to confirm the situation prior to notifying the NRC.
Unusual Event declared by licensee.
Subsequent'to the event, cemparison of an extrapolated cu ve for Reactor Vessel Head temper-ature with satur-ation temperature for RCS pressure indicates a steam void developed in the head at this time.
9:35 a.m.
NRC Senior Resident Insocetcr arrived in The SRI had been the Ginna Station Centrol Reem.
monitoring the ENS in his ci' ice since 9:33 a.m.
9:40 a.m.
Initial RCi pressure drop arrested at Termination of about 1138 psig.
pressure drep due to actiens of the SI pumps along with attaining i
saturation t
conditions in l
the Reactor Vessel i
Head.
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"B" Main Steam Isolation Valve manually closed and the "B" S/G was isolated.
Plant c;oling d:wn by dumping steam frem "A" S/G to the Mein Condenser.
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Time Event Comment _
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9:40 a.m.
Alert declared.
9:45 a.m.
Ginna Plant Superintendent notified the State of New York.
9:55 a.m.
NRC Region I Incident Response Center activated.
9:57 a.m.
Safety Injection initiation circuitry reset; containment isolation reset; instrument air restored to the centain-ment.
9:58 a.m.
Ginna Technical Support Center manned.
L10:04 a.m.
Charging pumps restarted.
10:07 a.m. (about) Pressuri:er PORY (430) manually cpened Shortly after the to reduce the pressure differential PORV was opened, between the RCS and ths "B" 5/G; Pressuri:er level Pressuri:er Relief Tank (PRT) temperature was sufficiently r
and pressure rise, high to cause the letcown orifice isolations and the inside containment letdown isolation (V-427) to open, resulting in lifting the let-down relief and acding water to the PRT.
10:08 a.m. (about) Pressuri:er PORV (4301 eanually cycled again.
l 10:03 a.m. (about) Pressuri:er PORV (430) manually opened Rapid rise in PZR and failed to shut; RCS pressure croppedc level without from about 1300 psig to about 900 psig; corresponding Pressuri:er level rises; PORV Block infection f1:w was Valve begins to close. Pressurizer level first clear indic-increased rapbJ1y.
ation in the Control Room that a steam void had formed in the Reactor Vessel Head.
The void was growing i
i as RCS pressure dropped.
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Time Event Ceneent 10:10 a.m. (about) PORV Block Valve closed; Pressurizer level goes offscale high; Safety Infection increases RCS pressure.
10:25a.m.(about)
"B" S/G atmospheric relief placed in manual control and closed in accordance with: procedure E-1.4 10:33 a.m. (about)
"B" S/G safety lifts (setpoint - 1085 Based on conversa-psig) and reseats. Safety Injection tions with Pumps secured to prevent further release operators.
through the "ft" safety.
All charging pumps are running.
10:42 a.m.
NRC Headquarters activated.
10:44 a.m.
Site Emergency declared.
10:4E a m. (about) PRT rupture disc ruptures, releasing Disc ruptured water to the "A" Containment Sumo, primarily due to PCRV (430) and the letdown relief with a minor contribution from the RCP seal return relief.
11:00 a.m.
Plant coo 1down now via *he "A" S/G Du= ping steam to atmospheric relief, the condenser secured to minimire.
spread of contamination in the secondary system.
11:15a.m.(about) One Safety Injection Pump restarted; Throttling of SI "B" S/G safety lifts and restats, based on informa-tion gained through Safety Injection throttled to prevent discussion with, further lifting of the "B" S/G safety, the licensee's Operations staff.
11:19 a.m.
The process computer fails. Remains out Licensee reading l
of service until 11:47 a.m.
incere and head thermoccur-r manualb to verify adequate core cooling.
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_ Event comment 11:20a.m.(about)
"A" RCP restarted; Reactor Vessel Head Time based on temperatures approach incore temperatures; graph of Incert the steam void 1,n the head collapses.
and Head therio-couples.
Data for this graph was obtained manually in the control room.
12:00 neon (about) Steam bubble drawn in the Pressuri:er.
Normal letdown r eestablished.
2:00 p.m.
Licensee reported Containment Sump A Channel 1 on at 9.3 feet (approx. 8000 gal.); PRT Containeent Sump at 92%.
indicated 5.3 feet (1900 gals.);
Channel 2 incicated 9.3
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Later, it was disc. overed that Channel 2 read incorrectly due to t static charge on the indicator.
4: 15p.m7 NRC Region 1 Incident Response Team onsite.
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C:40 p.m.
lavel in "S" S/G back within indicating range.
Plant cooling down by dumping steam from "A" 5/G to atmosphere.
A" RCP providing flow through the "A" loop and backflow through the "B" loop.
"B" 5/G being cooled by feeding the S/G with AFw' while blaecing it via the ruptured tube to the RCS.
7:17 p.m.
Site Emergency downgraded to Alert.
7.05 a.m., 1/25/32 RRR initiated to continue the cooldown.
"A" RCP remained in-operation.
10:45 a.m.
Alert downgraded to the Recovery Phase, 6:53 p.m.
Plant in Cold Shutdown.
5:30 p.m., 1/27/52 Containment Sump A pumped dry; total pumped - 1320 gallons.
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'ESPONSI JCS LETTER OF JANUARY 28. H I
Q Que aion 1.
What were the Comission's bases for allowin(Ginna to continue in operation in the f ace of the unresolved steavyEnerator problen and Ginna's history of steam generator tube f ailure?
Answer The NRC staf f has been evaluating steam generator operating experience on a case-by-case basis and for Ginna concluded that continued operation did not constitute an undue risk to the hesith or safety of the public.
This finding at Ginna has been based on'the followir.g considerations:
(1) Requirements for inservice inspections to monitor steam generator tube degradation have been established.
The f requency of inspection depends on previous adverse experience at Ginna.
Although the 151 program allows 24 months between steam generator tube inspections, Rochester Gas and Electric Corporation (RG&E) has inspected the Ginna plant on an average of every 6.6 months since 1972.
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(2) Adceptance criteria (plugging limits) have been established to ensure that degraded tubing will retain adequate structural mrgins over the full range of normal operating, transient, and postulated accident conditions.
(3) Should deradation develop completely (100%) through wall and leak, the resulting lea % age is generally small as indicated by operating experience.
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Allowaele limits on primary to secondary leakage (0.1 gallon per minute) have been established beyond which th! plant must be shut down for appropriate corrective action.
(4)
Information f rom operating experience and Unrtsolved Safety Issue (US1)
Action Plan efforts will be utilized to update interim criteria and requirements, (5) Wice dissemination of ALARA dose methods and techniques, based on up-to-date experience and further development ef forts, can help minimize total deses when steam generator inspection, repair, and replacenent are required.
Question 2:
What gJarantees can you offer the neighbors of plants afflicced with steam generator tube deterioration that these olants will not suf fer a similar accident, or one more serious?
Answer There is, of course, no absolute g'uarantee triat steam g(nerator tubes won't continue to dcteriorate at the Ginna f acility or at any other nuclear facility. 'We believe, however, that the required inspection program for steam ?,ene rat or tubee. Will keep the f requency of tube ruptures low and our ef f orts are cined at reducing this f reaucncy even Icaer. To c: ate there nave been four steam generator tube f ailures (gcetter than 50 gam) at pressurned water reators in the U.S.
The f acility, date of the event and estimated leakagt rate is as f ollows:
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Sorry Unit 2 07/15/76 80 Prairie Island Unit 1
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Ginna 01/28/82 700 1
The above data indicates that for all 48 PWRs licensed to operate in the U.S.
(as of February 1), about one tube f ailure has been occuring every two years since 1975. Preliminary information indicates that the leakage rate from the Ginna failure is approximately the maximum possible for a single tube failure; therefore, leakage mch in excess of this amount is not expected.
The second element of the NRC required review program includes the consideration
.of steam generator tube f ailures as design basis events; in f act sma!1 break j
loss-of-coolant accidents.
Plant emergency procedures and safety systems are
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designed to safely handle the complete range of postulated failures including
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a double-ended failure of the main coolant piping.
Therefore, it is expected that a plant could be brought to a safe shutdown following a tube failure in a 2 team generator with insignificant offsite j
consequences.
Question 4 What were the on and off-site doses from the Ginna accident, and what is the explanation for different figures reported to the press?
Answer Preliminary information regarding radioactive material releases as a result of the Ginna event is provided in the attached enclosure. This information is being investigated further and the results will be included in the 45-day Task Force report.
We have no information at this time of differing dose figures reported to the,
press..
L Question 4_.
Was there a leak before rupture at Ginna? If so, how does this alter your analysis of the risk of such an accident?
L Answer Although preliminary information does not indicate any leakage immediately prior to the stcan generator rupture event, this area is being further investigated and the results will be included in the 45-day Task F:rce report.
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e, Ques _ tion 5.
What actions is the NRC now taking to reduce the possibility of more such accidents?
Answer Our bases for licensing new plants as well as allowing continued operation of curreht plants is to assure that the steam generctors have and retain tube integrity without excessive leakage. To provide assurance that plants can be operated safely, the steam generators are tested initially to confirm tube integrity and plant Technical Specifications include requirenents for periodic inservice inspection of the tubes.
The Technical Specifications also include operating limits on prirary and secondary system activity levels. Tubes identified to be degraded beyond the limit specified in the plant Technical Specifications rust be removed from service by plugging.
For a few plants, repair f tubes by " sleeving" has been approved as an acceptable alternative to plugging thereby permitting the required tubes to remain in service.
In addition, the plant Technical Specifications provide limits on allowabic primary to secondary leakage, beyond which the unit must be shutdown f or additional inspection and repairs.
In addition, in 1979, the NRC established Unresolved Safety Issues A-3, A-4, and A-5 regarding degradation in W CE, and C&W steam cenerators, respectively. A draf t report, NUREG-Oe44, presenting the proposed NRC staf f resolution of these generic safety issues has been prepared and is currently under review.
The report integrates technical studies in the areas of systems analyses, inservice inspection (ISI), and tube intcgrity to establish improved criteria for ensuring adequate tube integrity and safe steam generator operation under all conditions.
The significant milestones for issuance of the final NUREG report are as follows:
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Issue Draft NUREG for Public Cornent 8/6/82 Receive Public Comments 10/1/82 Resolve Corments and Issue Final NUREG with Requirements 2/1/83 Further, the NRC has underway a confirmatory research program in the following areas:
(a) Steam Generator Tube Integrity (1 ) Establish burst and collapse pressures and leak rates for degraded tubes.
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(2) Efficiency of eddy current testing to locate and characterize defects in steam generator tubes.
(b) Stress Corrosion Cracking of Steam Generator Tubes Develop data and models to predict stress corrosion cracking service life of Inconel-600 steam generator tubes under normal and abnormal service conditions.
(c) Improved Eddy Current Inservice Inspection for Steam Generator Tubing Upgrade and validate eddy current inspection probes.
techniques, and associated instrumentation for inservice inspection of steam generator tubing.
Question 6.
What is the justification for issuing operating licenses to additional plants now nearing completion which contain similar steam generators?
Answer Our basis for licensing new plants is provided in response to Question 5.
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RELEASE POINT
_ ISOTOPE ACTIVITY, RELEASED Steam Jet Air Ejector Noble Gases 475 - 525 Ci I-131 0.001 - 0.002 Ci "B" Steam 4.enerator Hoble Gases 5 - 6 Ci I-131 0.015 - 0.075 Ci Mn-54 0.030 - 0.050 Ci Co-58 0.030 - 0.050 Ci Ba-140 0.17 - 0.30 Ci Note:
Short-lived isotopes not included.
Definitien of Curie (Ci): A unit of measure of the amount of radioactivity in a material.
One curie is equal to 37 billion disintegrations per second from the -nuclei of atoms.
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JANUARY 1932 THE ATTACMED SUF.%RY SHEET CONTAINS THE RESULTS OF THE TLD MONITO
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IN THE VICINITY OF GINMA FOR THE PERIOD WHICH INCLUDED THE' JANUARY 25,198?. INCIDENT.
THE COSE GIVEN IS THE GROSS ' DOSE MEASURED Te: 'T WITri NO CC'NTROL BADGE DOSE SUBTRACTED. THE ERROR GIVEN'IS A ONE-SID% STATISTICAL ERROR CNLY.
FOR COMPARISON, AN EXPECTED DOSE WAS CALCULATED USING DATA FOR THE FOURTH QUARTER OF 1981 AND
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NO DOSES WERE MEASURED WHICH WERE STAT'.STICALLY DIFFERENT FROM THE EXPECT'ED DOSEI.
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i DRAFT
$NOW SAMPLES, GINNA (micro-Curies / gram)
(104 C1/gm)
TRAIN _ING,, CENTER (ONSITE)
PUTNAM&FISHERRD.(OF7-SITE;-
ISOTOPE __
I-131 0.00009 0.0000005 1-133 0.00076 0.000004 Cs-137 0.00001 0.0000005 Cs-134 0.00007 0.0000003 Co-S8 0.00011 0.000003 Cr-$1 0.00006 0.000005 ANALYSIS BY NRC
~
I ISOTOPE
' TALLIES FIELD (NEAR SITE)RT. 104 & FISHER RD.(OFF SIT
( BOUNDARY)
~~
I-131 0.00001 4 0.0000001 7-133
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0.000009 0.0000003 Cr-51.
0,000006 4 0.0000007 NiALYSIS BY NRC ALL YALUES DECAY CORRECTED TO 9:26 a.m.,1/25/82 i
DPsAFT
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