ML17309A236

Forwards Petition for Order to Show Cause Why Facility OL Should Not Be Suspended or Why Permission to Restart Reactor Should Not Be Withheld Until Actions Have Been Taken to Assure Protection of Public Safety

ML17309A236
Person / Time
Site:	Ginna
Issue date:	03/11/1982
From:	Caplan R Sierra Club
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML17258A639	List: ML17258A637 ML17309A236
References
RTR-NUREG-0737, RTR-NUREG-737 NUDOCS 8203150125
Download: ML17309A236 (32)

Text

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ACCESSION NBR: 82031501?5 FACIL:50-244 BYNAME DOC ~ DATE: 82/03/11 NOTARIZED,' NO Robert Emmet Ginna Nuclear Planti Uni,t 1E 'Rochester

... G DOCKET 05000240 AUTH AUTHOR AFFILIATION CAPLANER AN ~ Sierra Club I

REC P ~ NAME RECIPIENT AFF IL I ATION DENTONEH ~ RE Office of Nu'clear Reactor Regulationp Director

SUBJECT:

Forwards petit)on for order to show cause why facility OL should not be suspended or why permission ',.to restart reactor should net be wiIthheld until, actions have 'been "taken

.to assure Ipr otection of >public safety.

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SIERRA -=:

CLUB 530 Bush Street San Francisco, California 94108 (415) 981-8634 Please reply to: 278 Washington Blvd.

Oswego, New York 13126 Harold Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Denton:

Enclosed for filing is a Petition For Ord se prepared by the Sierra Club. The petition pertains to the Ginna Nuclear Power Plant, Docket No. 50-244, and arises from the January 25, 1982, accident. As staff review of the accident is already in progress, we request prompt response to our petition.

Very tnuly yours, Ruth N. Caplan, Chair Sierra Club National Energy Committee Enclosure cc. with petition:

Senator Gary Hart Vawter Parker, SCLDF Senator Alan Simpson Joseph Fontaine, President, Sierra Club Congressman Morris Udall Eugene Coan, Sierra Club Congressman Richard Ottinger Jesse Riley, Nuclear Subcom, Sierra Club Congressman Edward Markey Richard Lippes, Chair, Atlantic Chapter Congressman Toby Moffett Beatrice Anderson, Chair, Rochester Group Richard Goldsmith, Esq. Robert Pollard, Union of Concerned Karin Sheldon, Esq. Scientists John E. Maier, Rochester Gas 8 Electric

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )

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Rochester Gas and Electric Corporation ) Docket No. 50-244

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R.E. Ginna Nuclear Power Plant )

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SIERRA CLUB PETITION FOR ORDER TO SHOW CAUSE INTRODUCTION This petition is broughtbefore the Office of Nuclear Reactor Regulation by the Sierra Club. Pursuant to 10 CFR 2.206, 50.54, 50.100 and 50.109, and for reasons set forth below, the Sierra Club requests Chat Rochester Gas and Electric Company be required to show cause, as provided in 10 CFR 2.202, why the operating license for Che Ginna nuclear reactor in Ontario, New York, should not be suspended, or in Che alternative, why permission Co re-start the reactor should not be withheld, until such time as essential actions have been taken by Che licensee and Che Commission to assure Che protection of public health and safety. The necessity for such actions arises from Che accident on January 25, 1982, which was initiated by a steam generator tube break and which triggered a site emergency.

In requesting this action, the Sierra Club wishes to stress our concern regarding Che potentially serious safety implications of the Ginna accident, not only to our 500 members living in Rochester, but also Co Che general public. Further, as a national environmental organisation with approximately 225,000 members across Che country and 18,000 members in New York State, we are concerned about the

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implications of Che Ginna accident for Che safe operation of other pressurized water reactors 1n New York and across Che country.

Given Che clear safety implications of both under- and over-pressurization which can arise subsequent; to a steam generator tube break, the Sierra Club concurs with Che November 24, 1981, "Informa-tion Report; Steam Generator Tube Experience" by NRC staff which states:

These Cubes, like many interface components, affect both Cprimary and secondary) systems, and their failure is an operational as well as a otential safet concern.

Therefore, Che steam generator must be viewed as part of Che total system in which it operates. Thus, maintaining Che integrity of Che tubes requires a systems approach that should encompass mechanical, structural, material, and chemical considerations. (page 35, emphasis added)

RELIEF REQUESTED The Sierra Club requests Chat Che Director of Nuclear Reactor Regulation in1tiate a full review by staff of matters pertaining Co the ability of the licensee to safely operate the reactor so as Co protect public health and safety, inight of the January 25th acci-dent. Such review should be made part of the review now in progress by staff and should include, but need not be limited Co, the specific areas detailed below. Pending completion of this review by the staff, the Operating License for Ginna should be suspended, or in the alter-native, re-start of Che reactor should not be permitted.

l. The cause of the Cube break initiating Che January 25, 1982, accident should be thoroughly explained and corrective action taken Co prevent such breaks in the future. The mechanical damage arising from loose pieces of metal should be studied in Che context of the generic corrosion problems at Ginna. Specifically, corrosion arising from AVT (all volatile treatment) control of secondary water chemistry should be addressed in relation to denting of tubes, stress

0 E I corrosion, and intergranular attack. This should include corrosion in Che feedwater system and corrosive impurities introduced by condenser leaks.

2. The adequacy of Che steam generator Cube testing program should be evaluated and a determination made regarding the following issues:

a. Is Che routine multi-frequency eddy current testing method being employed at Ginna Che best available given current state-of-Che-art? If not, what Justification is Chere for not employing Che best available technology, in light of chronic Cube degredation problems at Ginna and at other PWR's and Che existence of techniques such as fiber optic examination?

c. Does Che current testing program, which only tests a sample of Cubes and which does not test their full length, provide sufficient information to prevent tube failure?

3. The technical specifications defining the extent of allowable tube degredation for steam generator Cube rejections should be re-viewed in light of the Ginna accident to determine whether they are sufficiently stringent to prevent a Cube break.

4. The increased risk of steam generator Cube breaks/leaks, if RG8E operates the reactor without having proceeded with Che preventa-tive sleeving program originally scheduled for the Spring, 1982, refueling outage, should be assessed and a determination made as to whether the original schedule should be adhered to.

5. The safety implications of current and proposed plugging and sleeving of steam generator tubes and of further repairs such as insertion of stabilizing cables should be examined in order Co assess additional stress, such as from changes in fluid dynamics, which may

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be induced in tubes remaining in use.

6. An evaluation should be completed Co determine the safety implications of operator action currently required to re-establish the instrument air system and Co open the PORV manually.

7. The safety implications of Che failure of the PORV Co close should be assessed in light of the problems which developed during the Ginna accident, particularly with regard Co the, creation of a steam bubble in the reactor vessel as a result of depressurization. The potential for uncovering the core, due to a steam bubble in Che reactor vessel or elsewhere in Che primary system should be addressed.

A determination should be made as to whether safety funct1ons per-formed by the PORV require Chat it be designated as safety grade and be required 'Co meet all NRC regulations applicable to such safety grade designation, in order Co assure safe operation of the reactor.

8. A determination should be made, given the demonstrated unreliability of Che PORV, as to whether a reliable method exists for removing decay heat by means of Che secondary system, without providing, at the very minimum, one pathway for removing decay heat which consists of safety grade equipment. Such determinat1on should also include an assessment of Che reliability of essential auxiliary support systems such as instrument air, and should consider the con-sequences of loss of off-site power to determine whether General Design Criteria ¹17 of 10 CFR Part 50 Appendix A is met.

9. A determination should be made as to whether the emergency operator procedures set forth in "Westinghouse Emergency Operator Guidelines for Steam Generator Tube Rupture Events" are adequate to protect Che public health and safety. Operator delay, or apparent hesitancy, in terminating Che HPI (high pressure infection) is of particular concern in relation to Che risk of over-pressurization

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of the reactor pressure vessel as reported in Che Speis memorandum (see infra 811) and to the increased reliance on proper functioning I

of steam generator safety Valves. Further, Che Ginna emergency procedures should be conformed to Che Westinghouse guidelines.

10. The conditions under which the reactor vessel can become over-pressurized in Che course of operator action to contral an accident should be clearly specified and a determination made as to whether an automatic response system would decrease Che chance of over-pressurization problems from developing and,whether the instal-lation of such a system at Ginna is an action that "..Mill provide substantial, additional protection which is required for the public health and safety...." as provided in 10 CFR 50.109.

ll. The concerns raised in the Speis memorandum (Themis Speis to Roger Nattson, "Preliminary Evaluation of Operator Action for Ginna SG Tube Rupture Event" dated January 28, 1982, see infra Attachment E) regarding problems and potential problems in cooling the reactor following the Cube break should be addressed; a deter-'ination made as to their safety significance; and necessary corrective action taken. These include Che following problems:

a. the apparent stratification in the B steam generator and its effect on slowing depressurization of the faulted steam generator;

b. the consequence of an additional coolant system failure, including a leak in the A steam generator or'a secondary side safety/relief valve" sticking open;

c. Che necessity to remove decay heat from,.the A steam generator by steaming to the atmosphere due Co improper functioning of Che condensor;

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d. the problems associated w9;th the use of Che PORV for co'olant discharge during "feed and bleed" cooling.

12. A determination should be made as Co Che extent to which failure to implement the TMI Action Plan requirement for instrumenta-tion Co allow direct measurement of the water level in the reactor vessel contributed Co operator problems in determining proper timing for operating Che ECCS pumps and in determining Che size of the steam bubble.

13. A full investigation.;should be made to determine the state of embrittlemhnt of the Ginna reactor pressure vessel to determine the likelihood Chat over-pressurization will lead to vessel rupture as a consequence of pressurized thermal shock.

14. The NRC should determine whether Che reactor can operate safely without replacement of Che steam generator and associated parts of Che nuclear steam supply system and whether the newest Westinghouse steam generator design will ameliorate the problems, given Che recent problems which have developed with this design at McGuire and at t

European reactors.

15. The total pro)ected worker exposure should be calculated in advance of NRC approval of RGKE's repairs and a specific plan developed Co keep worker exposure as low as reasonably achievable (ALARA). This should include a determination as to whether time should be allowed for radioactive decay, particularly of Cobalt 58, in the steam genera-tor prior Co repairs, in order to prevent unnecessary worker exposure and still allow all necessary repairs to be made.

16. An overall safety assessment should be performed before Che reactor is allowed to re-start in order that the combined risk of potential failure modes can be determined, in relation to the protection of public health and safety. At a minimum such an assessment should

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address Che following:

a. Che degredation of the Ginna steam generators, including Che plugging, sleeving and other repairs required Co date and planned;

b. Che on-going contribution to tube degredation of corrosion arising from AVT control, from condenser leakage, and from Che feedwater system (as opposed to Che suspected damage from loose pieces of metal in the B steam generator);

c. Che lack of a safety grade pathway in the secondary system Co remove decay heat;

d. the chance that operator error will lead Co over- or under-pressurization of the reactor vessel;

e. the state of reactor vessel embrittlement.

The facts which constitute the basis for our request are set forth in Attachments A, B, C, D and E.

We respectfully request that a decision on our petition be rendered forthwith.

On behalf of Che Sierra Club, Respectfully submitted by, Ruth N. Caplan, Chair Sierra Club National Energy Committee 278 Washington Blvd.

Oswego, New York 13126 315-343-2412 I hereby affirm Chat the facts alleged herein are true and correct Co the best of my knowledge and belief.

DATED: March ll, 1982 Rut N. Caplan

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AFFIDAVIT OF BEATRICE ANDERSSN

1. My name is Beatrice Andersen. I live at 12 Spinet Drive, Rochester, Ginna New reactor York 14625, which owned by Rochester is about Gas and

~ miles from the Electric.

2. I am a member of the Sierra Club and I chair the Rochester Qroup of the Sierra Club which has ~50 members in the Rochester area.

3. On behalf of myself and the Rochester Group, I authorize the Sierra Club to represent my interests in the request for show cause action before the U.S. Nuclear Regulatory Commission.

These interests include the potential danger to my&health and safety if the Ginna reactor is allowed to restart prior to such actions as are called for in the Sierra Club show cause request;.

Sworn and subscribed to before me this day of ,1982.

EDWIN R. JEFFRIES JR.

otary Pubiic in the State of New York Notary Publ c MONROE COUNTY, NEW YORK Cornrnission Expires March 30, 19@ Z My commission expires

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ATTACHMENT . FACTUAL BASIS FOR SHOW SE PETITION

1. On January 25, 1982, a steam generator tube rupture at Che Ginna nuclear plant in Ontario, New York, occurred. The rupture occurred in a Cube which was last inspected in May, 1981, at which time the Cube showed less Chan 20$ was'~ of Che tube wall, according to "Weekly Information Report, February 18, 1982, from T.A. Rehn, Assistant for Operations Office of Che EDO Co Che Commissioners", included herein as Attachment B.

2. It is our understanding that RG&E has not yet been able to provide a satisfactory explanation for the rupture of the steam generator Cube.

Upon information and belief, a clear relationship has not been estab-lished between loose pieces of metal discovered in';the steam generator, the damaged peripheral tubes, and Che ruptured tube. An alternate explanation,linking the rupture to stress corrosion has been advanced by RG&E. (See Rehm memo, page 2 of Enclosure B)

3. Upon information and belief, the Ginna Cube testing program has been based on multi-frequency eddy current testing at Che time of refueling. Such testing has included only a sample of Cubes and only part of the tube length has been examined.. According to Nuclear Safet "most tubes were tested to the first support plate, some Co the sixth support plate, and a few over the U-bend." (Nuclear Safet p VS 22' 5p Sept.-Oct., 1981. Included infra as Attachment C.

4. Upon information and belief, the "Quality Assurance Manual, Ginna Station Inservice Inspection Program for the 1980-1989 Interval" allows the tube inspection interval to be extended to once every 40 months under certain conditions. Section 2.5 of this document states:

The inservice inspection intervals for the examination of steam generator tubes shall not be more than 24 months.

However, normal if fuel over a nominal two year period (e.g., Cwo cycles) at least two examinations of the separate legs result in less Chan 10$ of Che tubes with detectable wall penetration (O than 20$ ) and no significant (O,than 10$ ) further penetration of tubes with previous indications, the inspection interval of the individual legs may be extended Co once every 40 months. (page 5 of 22)

5. Upon information and belief, RGRE reported Co the NRC staff on February 10, 1982, Chat tests after the accident did not reveal serious problems with Che steam generator Cubes which would prevent RG&E from re-starting the reactor. Yet After fiber optic examination was required by staff, serious problems were found in tubes previously plugged.

John Maier, RGRE Vice-president for Electric and Steam Generation, commented to the press the next day: "The pictures are very dramatic....

It looks like somebody went in with a hacksaw. Some of the Cubes show (AP quot;ed in Palladium-Times, severe denting and external degredation."

Feb. 12, 1982) Further examination revealed Cwo pieces of metal weighing "'a couple of pounds'...with one of Chem as large as 6.5 x 4 inches and seven-sixteenths inches thick." (Nucleonics Week Feb. 18, 1982 As reported in Nucleonics Week, Feb. 25, 1872, one RGRE source stated:

"'Some are corroded, some are imploded, some are gust sheared.'"

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6. Upon information and belief, RG&E was planning an extensive sleeving program Co remedy corrosion problems regarding Che steam generator tubes.

In a letter from John Maier Co Dennis Crutchfield, January 15, 1982, RG8E requested permission to "delete Che 25 sleeve limitation" so Chat more sleeves could be installed during each steam generator inspection.

(See infra, Attachment D.)

7. As recently as September 21, 1981, Ginna was not listed as one of the 11 units with the most serious steam generator problems (New York Times, Sept. 21, 1981, B-10). It is our opinion Chat this fact emphasizes the unpredictable nature of Che rupture and reinforces Che need for much more stringent test procedures.

8. Upon information and belief, Che introduction of AVT control of secondary water chemistry at Ginna has led Co problems of intergranular attack and tube corrosion, requiring Che plugging of steam generator tubes. (Nuclear Safet , Ibid.)

9. As indicated in the Point Beach proceedings, AVT control does not function to precipitate out solid impurities that leak into Che generator and does not prevent build-up of hardness scale on the heat transfer surfaces. Both conditions degrade steam generator tubes.

(Docket 6630, ER-10, Exhibit 16E at 14-15)

10. As observed by NRC staff, "denting" of steam generator tubes oacur-red in several PWR facilities, including Turkey Point, Units 3 and 4, and Surry, Units 1 and 2, after 4 Co 14 months of operation, following Che conversion from a sodium phophate treatment Co AVT chemistry for the steam generator secondary coolant. (" Information Report Steam Generator Tube Experience, November 24, 1981, SECY 81-664,"'Appendix B, page 3.) We note hte report's observation Chat: "Tube denting is most severe in Che rigid regions or so-called 'hard spots'n Che tube support plates. These hard spots are located...around Che peripheral locations of Che support plate where Che plate is wedged to Che wrapper and shell." (Ibid., page 3) Upon information and belief, the staff has already requested Chat RGRE have Westinghouse prepare a report regarding this matter.

11. The NRC "Information Report Steam Generator Tube Experience" con-cludes: "copper alloys should be eliminated from all areas of Che condensate/feedwater/steam condensation cycle. Substantial evidence exists Chat copper oxides in Che steam generators are an important catalyst in accelerating the rate of corrosion processes within the steam generator s . " (Ibid-., p. 4 2 )

12. Condens'er leakage is also relevant Co the action at hand. Staff states: "With the exception of a few reactors which are sited where no acid producing species exists in Che condenser cooling water, all currently operating plants are susceptible to denting, if condenser leakage occurs. Because copper oxide has been demonstrated sufficient to be a catalyst, those plants with copper in"Chbir secondary cycles are even more susceptible." (Ibid., Appendix A, page 6)

13. Steam generator problems are not automatically solved by installing new steam generators as evidenced by Che problems faced by Prairie Island 2 and by North Anna 1. Brookhav~ National Laboratory commented

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I'p<>hh 'Il f I>> P g t, I. Vlkf <<1>>e, ~ 1 W l ~ ~ '<< ~ ~ l'P 'I Ie>> 1 'hh I ttr I'I 1 bl I 1 1- t VVI 't ~ 1 " r P I T t' bf Ib>> ~ I ' ~ ' hb P >>h ~ J '1>> 'gt >> $~ P ~ P I Attachment A, page ' 3 ~ ~ last year as follows: It seems ironical that Prairie Island 2, which has no'i copper in the system, stainless steel condensers, and meticulous monitoring of water chemistry, should be the one unit Co have suffered from this particular phenomenon (of Cube corrosion): Che Prairie Island Units have to date been a~. ihind.ng~<examp3.e of what we thought was Che proper way Co avoid corrosion problems;-'-', I,- (Docket 6630, CE-20, Exhibit 40,. p.3) Such experiences make it all Che more imperative to have a stringent testing schedule for tubes and strict standards for removing tubes from service.

14. Upon information and belief, Che sequence of events during the January 25 accident clearly indicate Che interdependency of the nuclear steam supply system and the reactor safety system. Reactor trip in response Co Che tube break initiated containment isolation which resulted in loss of instrument air. This required operator action to open Che PORV manually, when Che valve was required Co relieve over-pressurization. The reactor vessel became under-pressurized when the PORV stuck open and the block valve had to be closed. Lowered pressure produced a steam bubble in the top of Che reactor vessel when water flashed to steam. A second drop in pressure about 30 min-utes later again led Co water in Che reactor vessel flashing Co steam.

(Source: "Preliminary Evaluation of Operator Actions for Ginna SG Tube Rupture Event" by Themis Speis. See infra Attachment E.)

15. Upon information and belief, Che Speis memo. also indicates Chat over-pressurization of the reactor vessel was of concern during the sequence of events during which operators Cried to stabilize Che reactor. First, charging pumps were restarted before Che B steam generator was isolated, leading Co a build-up of reactor pressure.

Second, Che SI pump was restarted without apparent need Co do so, which has elicited concern regarding operator hesitance to terminate HPI and Che consequence for pressurized thermal shock.

16. According Co the "Information Report Steam Generator Tube Exper-ience," the total man-rems exposure can be quite significant. The report states: "Where ma)or repair or replacement efforts are re-quired, dose expenditures may range from 2000 Co 3500 man-rems." (Ibid, page 51) The largest dosage reported results from steam generator repair at San Onofre Unit 1, where 3493 man-rems exposure is reported for the 273-day outage during 1980-3.981. (Ibid, Table 6) This is more than Che 1759 man-rems for steam generator replacement at Surry, Unit 1 or the 2140 man-rems for Surry, Unit 2 replacement. (Ibid., Appendix B, page 13 and Table 6) It is our belief Chat these dose levels point Co Che need to evaluate total man-rems exposure in determining the best course of action to be followed at Ginna.

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~ ~ 4 ~ ~- . 1 Kv II ~ 4 ~ g 4 ~ I If' t 1 ~ I , IK v J ~,I 4 4>>, (4 ~ 1 h ~ 1 II ,4 rh ~ V 4 I>> ~ I ~ ACHNENT B Februar 18, 1982 For: The Commissioners From: T. A. Rehm, Assistant for Operations, Office of the EDO ~Sub'ett: WEEKLY INFORMATION REPORT - MEEK ENDING FEBRUARY 12, 1992 A su+nary of key events is included as a convenience to those Commissioners who may prefer a condensed version of this report. Contents Enclosure Administration A Nuclear Reactor Regulation Nuclear Material Safety and Safeguards Inspection and Enforcement Nuclear Regulatory Research . Executive legal Director International Programs G State Programs Management and Program Analysis Controller Analysis and Evaluation of Operational Data K Small S. Disadvantaged Business Utilization L Regions Items Approved by the Commission T. A. Rehm, Assistant ~ ~ for Operations Office of the Executive Director

• No input this week. for Operations

Contact:

T.. A. Rehm, OEDO 49-27781 fOR SUBSCRIBERS ONLY

R. E. GINNA The ruptured tube in the Ginna steam generator was inspected in May 1981.

The ECT results showed that there was 20K penetration (an OD signal) 3 to.,

6 in. above the tubesheet. The failed tube is in row 42, column 55 which is near the periphery of steam generator. It is located in the "wedge area" of the steam generator. This is the section of the support plates '..

that, does not have floe holes. Three of the six previous small leaks .

that have been experienced have been in that'r ea: There is no sludge pi1e in that area.

2. The ruptured tube in the Ginna steam generator has been inspected .using fiber optics. The rupture has been determined to start approximately 2, to 3 inches above the tube sheet and is approximately 5 inches long.

The rupture is -kite-shaped with a maximum width of 3/4 to 7/8 inch.

ostulates that the ru ture was due to stress corrosion li esto

'G&E differential ex ansion between the tube and the tube wra per in the ed e re ion (a region. where the tube support plate is fastened to the '

wrapper . Profilometry, to determine bulging or unusual shape of the tubes, showed some bowing of the tubes in the area of the rupture, thus adding credance to this theory. The ruptured tube is being plugged and removal of the tube is not anticipated'due to its location in the tube bundle.'GEE is planning to use fiber optics to inspect the ruptured tube from the secondary side .

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Eddy current testing (ECT) of the "B" steam generator has been completed.'n addition to the rupture'd tube, twenty other hot leg tubes are scheduled to be plugged.. Three of the tubes are adjacent to the ruptured tube while the others are tubes nonrelated to the accident that indicate intergranular attack (IGA) or >405 degradation. ;No plugging, other than the ruptured tube, is planned for the cold leg of the "B" steam generator.'GEE has commited to ECT 100% of the "A" .steat'enerator hot leg tubes plus all periphery tubes and a random sample of the sludge area tubes in the cold leg of the "A" steam generator.

ENCLOSURE B

R. E. GIHHA Cont'd

3. On Wednesday, February 10, 1982, members of the HRC staff met with representatives of Rochester Gas and Electric Corporation (RGB) to discuss the requirements to be met prior to restart of the R. E. Ginna Nuclear Power Plant. RGAE had scheduled the startup of Ginna for Monday, February 15, 1982, and proposed operation until the scheduled May 15, 1982 refueling outage, at which time the plant would be shut

. down and eddy current tests (ECT) 'of'he steam generators (S/G) would be performed. RGEE presented information on the cause and corrective action for the tube located in the wedge area that ruptured. In addi-

.tion, there was a description of Power Operated Relief Yalve (PORY) modifications and discussion of emergency procedures.

'-RGhE'has performed extensive ECT of both S/Gs. The ruptured tube has been inspected using. fiber optic equipment and a videotape of the rupture was sho~n at the meeting. Fiber optic inspection of the secondary side of the "B" S/G is in progress.'n addition.to the failed tube, the

. licensee has plugged 20 additional tubes in the "B" S/G because of inter-granular attack or wastage indications.

The staff has concluded that there was not sufficient technical basis resented at this time to permit the Ginna plant to return to operation.

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pecifically, the staff felt that prior to restart RGAE should:

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'1. Finish the fiber optic inspection of the secondary side of "B" 5/G including inspection for loose parts;

'Z. Obtain the S/G designer's opinion on the effects of plugging in the wedge area; and

3. Provide a more complete basis for operating for the proposed 3 months.

.A meeting to discuss the remaining'areas will be set up when RGB has prepared their response.

I EHCLOSURE B

l ATTACHMENT C Fariey 1, USA Denting, a phenomenon caused by ingress of cltloride leading to acid. forming conditions, results in One leahng tube was plugged at Farley 1. Thc nonprotcctivc corrosion product deposition in tube-to-defect was located at the U.bend, but the cause of tubc.support annuli in stcam generators with drilled-failure was not determined.

hole carbon steel support plates. It has been postulated Eddyeurrent inspection was performed on 153 that the addition of boric acid to secondary water tubes in steam gcncrator A and 306 tubes in stcam mitigates denting by forming stable, protective iron generator C, where the leaking tube was located. borates. This treatment is now being used at Indian Remote television inspection was used to augment Point 2.

a eddyeurrent testing, Of the 437 tubes plugged in thc four steam generators at Indian Point 3, denting defects were Ginna, USA observed'in 69 tubes at support plate intersections.

Because denting causes inward distortion at the sup-Nineteen tubes, all in steam generator B, werc plugged at Ginna during 1979. Tlurtecn of the tubes port plate, giving rise to the potential for SCC at thc had indications of intergranular attack in the tube small-radius U-bends, all tubes in row I were plugged sheet crevice; two tubes showed wall thinning just (368 tubes).

above the tube sheet. Tube corrosion by intergranular The stcam generators at Indian Point 3 were caustic SCC is typical of steam generators with a long inspected by techniques commonly used at plants with tube sheet crevice. At Ginna, these failures have significant denting. This includes using eddywurrent occurred every year since 1975, the year after intro. probes of different diameter and photographing thc duction of AVT control of secondary-water chemistry. secondary side to measure distortion of flow slots. Thc The wall thinning at support plates 1 and 2 was sludge deposit on the tube sheet was found to be soft, thought to be caused by water flashing to steam in the and'it was estimated that M2% could be removed by annulus during the early years of operation. These lancing with water. Boric acid is added to steam annuli are now packed with corrosion products. Other generators during condenser leakage.

tubes have this type of defect, but the thinning is

<20% of the tube wall thickness. The wastage defects Jose Cabrera, Spain are thought to be caused by a hydraulic-mechanical Three tubes were plugged because of fret ting at the mechanism rather than corrosion because all affected antivibration bars, and one was plugged because of tubes are in the periphery of the buridle where sludge phosphate wastage just above the tube sheet. Only does not normally accumulate. seven tubes have been plugged in the Jose Cabrera a

Tube inspection was performed by mtdttfrequen~c steam generator in 2915 EFPD of operation with eddy.current testing, as in 1977 arid tpyg. The phosphate treatment of secondary water, and six of inspection pattern was sim11ar to that of t97g: most these failures were caused by fretting at the antivibra-tubes were tested to. the first support plate, some to tion bars.

the sixth support plate, and a few over the.Upend.. Multifrequency eddyeurrent testing was used to About 2000 tubes were tested in each steam generator, inspect 80 tubes at the U-bend and almost all tubes to with a 5: 1 ratio between the stot and cold legs. the first support plate. Phosphate wastage of 40 to 49%

Ginna was the first PNR station with recirculating of the tube wall was detected in six tubes (including 1 steam generators to use Ml-flow deep-bed condensate that was plugged), and wastage of 30 to 39% was demineralization in the United States.v Very good detected in 46 tubes. This is the first reported instance experience has been reported with steam generator of phosphate wastage at Josd Cabrera.

water chemistry control and with the operation of the demineralizer system. KKS Stade, Federal Republic of Germany e

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Eddy<urrent. Inspection. of 574 tubes in steam Indian Point 2 and 3, USA generator I and 1262 tubes in steam generator 2 showed that three tubes in steam generator 1 and 56 in Twenty-six steam generator tubes were plugged at steam generator 2 had phosphate wastage of <25% of Indian Point 2 because of reduced tube diameter at the the tube wall. Two tubes were removed for metallur-support plates. Thcsc defects were found by cddy- gical examination. Stade, like Borssele, has Incoloy 800 current inspection of 1519 hot-leg tubes. tubes and has used low-phosphate treatment (2 to 6 mg NUCLEAR SAFETY, Vol. 22, No. 5, September-October 1981