ML17258A637
| ML17258A637 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/11/1982 |
| From: | Caplan R Sierra Club |
| To: | NRC COMMISSION (OCM) |
| Shared Package | |
| ML17258A639 | List: |
| References | |
| NUDOCS 8203150127 | |
| Download: ML17258A637 (12) | |
Text
UNITED STATES OR AMERICA NUCLEAR REGULATORY COMMISSION In Che Matter of))Rochester Gas and Electr1c Corporation
))R.E.Ginna Nuclear Power Plant)Docket No.50<<244 SIERRA CLUB PETITION FOR ORDER TO SHOV CAUSE INTRODUCTION This petition is broughtbefore the Office of Nuclear Reactor Regulation by the Sierra Club.Pursuant to 10 CRR 2.206., 50.54, 50.100 and 50.109, and for reasons set forth below, the S1erra Club requests Chat Rochester Gas and Electric Company be.required to show cause, as provided in 10 CPR 2.202, why Che operating license for the Ginna nuclear reactor in Ontario, New York, should not be suspended, or in Che alternative, why permission to re-start Che reactor should not be withheld, until such time as essential actions have been taken by the licensee and the Commission to assure the protection of public health and safety.The necessity for such actions arises from the accident on January 25, 1982, which was 1nitiated 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 the potentially serious safety 1mplications of the Ginna accident, not only to our 500 members living in Rochester., but also Co the general public.Further, as a national environmental organization with approximately 225,000 members across Che country and 18,000 members 1n New York State, we are concerned about the 8203 2 PDR~OR WOCCW'~OOOŽn~
g l t4" implicat1ons of the Ginna accident for the safe operation of other pressurized water reactors in New York and across the 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, l98l,"Xnforma-tion Report-Steam Generator Tube Experience" by NRC staff which s6ates: These tubes, like many Interface components, affect both Cprimary and secondary3 systems, and their faIlure is, an operational as well as a otential safet concern.Therefore, the steam generator must be viewed as part of Che total system In which it operates.Thus, maintaining the integrity of the Cubes requires a systems approach that should encompass mechanical, structural, material, and chemical considerations.(page 35, emphasis added)RELIEF REQUESTED The Sierra Club requests that the Director of Nuclear Reactor Regulation Initiate a full review by staff of matters pertaining to the ability of Che licens'ee Co safely operate.Che reactor so.as to protect public health and safety, In lighbof Che 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, Che 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 the reactor should not be pe'rmitted.
\l.The cause of the tube break initiating the January 25, 1982,'ls accident should be thoroughly explained and corrective action taken Co prevent such breaks in Che future.The mechanical damage arising from loose pieces of metal should be studied in the 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 c~<'
corros1on, and intergranular attack.This should Include corros1on in the feedwater system and corrosive 1mpurities introduced by condenser leaks.2.The adequacy of the steam generator tube testing program should be evaluated and a determination made regarding the follow1ng issues: a.Is the routine multi-frequency eddy current test1ng method being employed'at Ginna the best available given current state-of-the-art2 If not, what Justification is there for not employing the best available technology, 1n light of chronic tube degredation problems at G1nna and at other PWR's and the existence of techniques such as fiber optIc examination'?
c.Does the current test1ng program, which only tests a sample of tubes and which does not test their full length, provide suff1cient informat1on to prevent tube failure?3.The technical specificat1ons defining the extent of allowable tube degredation for steam generator tube regections should be re-viewed in light of the Ginna accident to determine whether they are suffic1ently stringent to prevent a tube break.4.The increased risk of steam generator tube breaks/leaks, if RGRE operates the reactor without having proceed'ed with the 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 adhe ed to.5.The safety implicat1ons of cur ent and proposed plugging and sleeving of steam generator'ubes and of further repairs such as insert1on of stabilizing cables should be examIned in order to assess additional stress, such as from changes in fluid dynamics, which~;n=':g'"
0'L be induced 1n tubes remaining in use.6.An evaluation should be completed to determine the safety 1mplications of operator action currently required to re-establish the instrument air system and to open the PORV manually..
7.The safety implications of the failure of the PORV to close~~should be assessed in light of the problems which developed during psC/R>8 the G1nna accident, particularly with'regard to the.creation of.a steam bubble in the reactor vessel as a result of depressurization.
The potent1al for uncovering the'core, due to a steam bubble in the reactor vessel or elsewhere in the primary system should be addressed.
A determination should be made as to whether safety functions per-formed by the PORV require that it be designated as safety grade and be required to meet all NRC regulations applicable to su'ch safety grade designation, in order to assure safe operation of the'reactor.
8.A determination should be made, given the demonstrated unreliab111ty of the PORV, as to whether a reliable method exists for removing decay heat by means of the secondary system, without prov1d1ng, at the very min1mum, one pathway for removing decay heat which consists of safety grade equipment, Such determinat1on should also include an assessment of the reliab111ty of essential auxiliary support systems such as instrument air, and should consider the con-sequences of loss of off-site powe to determine whethe General ((C g,(O'esign Criteria 817 o 10 CFR Part 50 Appendix A is met.9..A determination should be made as to whether the emergency ope ator procedures set forth in"'rlestinghouse Emergency Operator Guidelines for Steam Generator Tube Rupture Events" are adequate to protect the public health and safety.Operator delay, or apparent hesitancy, in terminating the HPZ (high pressure infection) is of part1cular
'concern in relation to the risk of over-pressurization I ,~of the reactor pressure vessel as reported in the Speis memorandum (see infra$11)and to the increased reliance on proper functioning of steam generator safety Valves.Further, the Ginna emergency procedures should be conformed'o the westinghouse guidelines.
10.The conditions under which the reactor vessel can become over-pressurized in the 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 the 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 50el09.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 potent1al problems in cooling the reactor following the tube break should be.addressed; a deter-mination made as to their safety significance; and necessary correctiv action taken.These include the follow1ng problems: a.the apparent stratification in the B steam generator and its effect on slowing depressurization of the faulted steam generator; j'b.the conseauence of an additional coolant system failure, including a leak in the A steam generator or"a secondary side safety/relief valve" sticking open;s c.the nec'essity to remove decay heat from the A steam gene ator by steaming to the atmosphere due to improper functioning of the condensor;
~I 8, d.the problems associated with the use of the PORV for y+(P coblant discharge during"feed and bleed" cooling.12.A determination should be made as to the extent to which failure to Implement the TNX Action Plan requirement for Instrumenta-tion to allow direct measurement of the water level in the reactor vessel.contributed to operator problems, 1n determining proper timing for operatIng the ECCS pumps and In determining the size of the steam hubble.13.A full investigation should be made to determine the state, of embrittlemhnt of the Ginna reactor pressure vessel to determine~pP the likelihood that over-pressurizatIon will lead to vessel rupture as a consequence of pressurIzed thermal shock.14.The NRC should determine whether the reactor can operate safely without replacement of the steam generator and assocIated parts of the nuclear steam.supp'ly system and'hether the newest Westinghouse steam generator design will ameliorate the problems, given the recent problems which have developed wIth this design at NcGuire and at European reactors.15.The total pro)ected worker exposure should be calculated in advance of NRC approval of RGRE's repairs and a specific plan develope to keep worke exposure as low as reasonably achievable
{ALARA).This should include a determination as to whethe time should be allowed for radioact"ve decay, particularly of Cobalt 58, in the steam genera-/tor prior to repairs, in order to'prevent unnecessary worker exoosure and still allow all necessary repa'rs to be made.16.An overall safety assessment should be pe formed before the reactor Is allowed to re-start in order tha" the combined risk o potential failure modes can be determined, in relation to the protection of public health and safety.At a minimum such an assessmen ent should I I address the following:
a.the degredation of the Ginna steam generators, 1ncluding the plugging, sleeving and other repairs requ1red to date and planned;b.the on-going contr1bution to tube degredat1on of corrosion arising from AVT control, from condenser leakage, and from the feedwater system (as opposed to the suspected damage from loose.pieces of.metal in.the B steam generator);
c.the lack of a safety grade pathway in the secondary system to 5<Jc.8'.remove decay heat;d.the chance that operator error will lead to over-or under-J)5>$8 pressurization of the reactor vessel;e.the state of reactor vessel embrittlement.
The facts which constitute the bas1s for our request are set forth in Attachments A, 8, C, D and E.Me respectfully request that a decision on our petition be rendered forthwith.
On behalf'f the SIerra Club, Respectfully submitted by, Ruth N.Caplan, Chair Sierra Club National Energy Committee 278 washington Blvd.Oswego, New'ork 13126 315-343-2412 I hereby affi m that the facts alleged he e1n ar true and correct to the best of my knowledge and belief.DATED: March ll, 1982 Rut J.Caplan AFFIDAVIT OF BEATRICE ANDERSSN l.My name is Beatrice Anderson.I live at 12 Spinet Drive, Rochester, New York 14625, which is about~miles from the Ginna reactor owned by Rochester Gas and Electric.2.I am a member of the Sierra Club and I chair the Rochester Group of bhe Sierra Club which has~$0 members in bhe Rochester area.3.On behalf of myself and the Rochester Group, I authorise the Sierra Club to represent my Interests in the request for show cause action before the U.S.Nuclear Regulatory Commission.
These interests include the potential danger to my:.health and safety if the Ginna reactor is allowed to restart prior to such actions as are called for in the Sierra Club show cause request.Sworn and subscribed to before me this~re~day of ,l982.rEP~RIES JR,'.CterY posEc in the State".f Near York-~MONROE COt.'I'llY, iGV YORK Commission Eroires March 30, 14@w My commission expires
~'0 r L ATTACHMENT A.FACTUAL BASIS FOR SHOW CAKE PETITION 1.On January 25, 2.982, a steam generator tube rupture at the GInna nuclear plant In Ontario, New York, occurred..The rupture occurred in a tube which was last Inspected In May, 1981, at which time the tube showed less than 20$wa~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 Che rupture of the steam generator tube.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 Cubes, and the 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, Che Ginna Cube testing program has been based on multi-frequency eddy current testing at the time of refueling.
Such testing has Included only a sample of Cubes and only part of Che tube length has been examined.According Co Nuclear Safet , Fmost tubes were tested Co the f'irst support plate, some to the sixth support plate, and a few over the U-bend." (Nuclear Safet , V.22, N.5, Sept.-Oct., 1981.Included infra as Attachment O.Upon Information and belief, the"Quality Assurance Manual, GInna Station-Inservice Inspection Program for the 1980-1989 Interval" allows Che tube Inspection interval Co be extanded Co 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, if over a nominal two year perIod (e.g., two normal fuel cycles)at least two examinations of Che separate legs result in less Chan 10$of the tubes with detectable wall penetration (W Chan 20%)and no significant (0 than 10~~)further penetration of Cubes with previous Indications, the InspecCIon Interval of the individual legs may be extended Co once every 40 months.(page 5 of 22)5.Upon information and belief, RG&E reported to the NRC staf f on Februarv 10, 1982, that tests after the accident.did not reveal se Ious problems with Che steam generato tubes which would prevent RG&E from re-starting the reactor.Yet After fiber optic'xamination was required by staff, serious problems were found In tubes previously plugged..John Maier, RG8E 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 tubes show severe denting and external dearedation." (AP quoted In Palladium-.imes, Peb.12, 1982)Further examination revealed two pieces of mezal weighing"'a couple of pounds'...with one of Chem as large as 6.5 x 4.Inches and seven-sixteenths Inches thIck."{Nucleonics Week'eb.18, 1982)As reported In Nucleonics Week, Peb.25, 1872, one RG&E source stated: "'Some are corroded, some are Imploded, some are, fust sheared.'"~\'
Attachment A.page 2~I~~~z 6.Upon Information and belief, RGhE was planning an extensive sleeving program Co remedy corrosion problems regarding the steam generator tubes.In'a letter from John Maier to Dennis Crutchfield, January 15, 1982, RGEE requested permission Co"delete the 25 sleeve limitation" so that 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.Che 11 units with Che most serious steam generator problems (New York Times, Sept.21, 1981, S-10).It is ouz opinion that this fact emphasizes the unpredictable nature of the ruptuz'e and reinforces the need for much more stringent test procedures.
8.Upon information and belief, the 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 Che Point Beach proceedings, AVT control does not function to precipitate out so11d Impurities that leak into the 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 Cubes occur-red in several PWR facilities, including Turkey Point, Units 3 and 4, and Surry, Units 1 and 2, after 4 to 14 months of operatIon, following Che conversion from a sodium phophate treatment to AVT chemIstry for the steam generator secondary coolant.(" Information Report-Steam Generator Tube Experience, November 24, 1981, SECX 81-664,"Appendix B, page 3.)We note hte report!s observation that:."Tube denting is most severe In the rigid regions or so-called'hard spots'n the tube support plates.These hard spots are located...around the peripheral locations of the support plate where the plate is wedged to the wrapper and shell." (Ibid., page 3)Upon information and belief, the staff has already requested that RGAE have Vestinghouse prepare a report z egarding this matter.11.The NRC"Information Report-Steam Generator Tube Experience" con-cludes: "copper alloys should be eliminated from all areas of the condensate/feedwater/steam condensation cycle.Substantial evidence exists that copper oxides in Che steam generators are an important catalyst in accelerating the rate o corrosion processes w" thin the steam generators." (Ibid., p.22)12.Conden.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 the condenser'cooling water, all currently ope ating plants are susceptible to dentIng, If su¹icient condenser leakage occurs.Because copper.oxide has been demonstrated to be a catalyst, those plants with cooper in their.secondary cycles are even more susceptible." (Ibid., Apoendix A, page 6)13.Steam generator problems are not automatically solved by installing new steam generators as evidenced by the problems faced by Prairie Island 2 and by North Anna 1.Brookhav~National Laboratory commented Attaehmynt A, page 3~~~e C 4 1 N~t1'Le~AL','t AJ LA e V~M O'Itt~A I~4 last year as follows: It seems ironical that Prairie Island 2, which has no copper in the system, stainless steel condensers, and meticulous monitoring of water chemistry, should be Che one unit to have suffered from this particular ph'enomenon (of tube corrosion):
Che Prairie Island Units have to date been a..;shining.
example,e of what we thought was the proper way to avoid corrosion problems.(Docket 6630, CE-20, Exhibit 40, p.3)Such experiences make it all Che more.imperative to.have a stringent testing schedule for Cubes and strict'standards for remov1ng tubes from serv1ce.14.Upon 1nformation and bel1ef, Che sequence of events dur1ng the January 25 accident clearly indicate the interdependency of the nuclear steam supply system and the reactor safety system.Reactor Cr1p in response Co the tube break 1nit1ated containment 1solat1on wh1ch resulted in loss of 1nstrument a1r.This required operator action to open the PORV manually, when Che valve was requ1red to relieve over-pressurization.
The reactor vessel became under-pressurized when Che PORV stuck op'n and the block valve had Co be closed.Lowered pressure produced a steam bubble in Che top of Che reactor vessel when water flashed to steam.A second drop in pressure about 30 min-utes later again led to water in Che reactor vessel flashing to steam.(Source: "Preliminary Evaluation of Operator Actions for Ginna SG Tube Rupture Event" by Themis Speis.See infra Attachment E.)I 15: Upon information and belief, the Speis memo also indicates that over-pressurization of the reactor vessel was of concern dur1ng the sequence of events during wh1ch operators.Cried to stabilize the.reactor.First, charging pumps were restarted before Che B steam generator was isolated, leading to a build-up of reactor pressure.Second, the SI pump was restarted without apparent need to do so, which has elicited concern regarding operator hesitance Co term1nate HPI and Che consequence for pressurized thermal shock.16.According to the"Information Report-Steam Generator Tube Exper-ience," the total man-rems exposure can be quite significant.
The report states: "'r/here ma)or repair or replacement efforts are re-quired, dose expenditures may range from 2000 Co"3500 man-rems." (Ibid, pane 51)The largest dosage reported results from steat'enerator repair at San Onofre Unit l, where 3493 man-r ms exposure is reported for the 273-day outage during 1980-1981.(Ibid, Table 6)This is more than the 1759 man-r ms for steam generator replacement at Sur y, Unit 1 or Che 2140 man-rems for Surry, Unit 2 replacement.(Ibid , Appendix"B, page 13 and Table 6)It is our belief that these dose levels@oint to the need to evaluate total man-rems exposure 1n determining the best course o action to be followed at Ginna.