ML20203E342
| ML20203E342 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 02/11/1998 |
| From: | Collins S NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20203E098 | List: |
| References | |
| 2.206, DD-98-01, DD-98-1, NUDOCS 9802270028 | |
| Download: ML20203E342 (4) | |
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7590 01 P UNITED STATES NUCI KAR REGULATORY COMMISSION NORTHEAST UTILITIES MILLSTONE NUCI FAR POWER STATION. UNITS 1. 2. AND 3 DOCKET NDS. 50-245. 50 336. AND 50-423 j:nQQQA NECK PLANT DOCKET ND. 56213 ISSUANCE OF DIRECTOR'S DECISION UNDER 10 CFR 2 208 Notice is hereby given that the Director, Office of Nuclear Reactor Regulation, has issued a directors decision with regard to a Petition dated March 3,1997, filed by Mr. Albert A. Cizek, i
l
! aroinafter refensd to as " Petitioner." The Petition pertains to the Millstone Nuclear Power i
Station, Units 1,2, and 3, and the Haddam Neck Plant.
i The Petitioner requested that the NRC:
1.
Within 30 calendar days of receiving a total of three license violations from the U.S.
- ~
Nuclear Regulatory Commission [NRC) duing any [3-year) period, irrespective of the violation level, the operating license of the facility shall be suspended for a period of not less than 90 days and not more than 180 days, 2.
Within 30 calendar days of receiving a total of three violations, of 10 CFR Part 50, including all applicable appendices, from the [NRC] during any [3 year] period, irrespective of the violation level, the operating license of the facility shall be i
suspended for a period of not less than 90 days a7d not more than 180 days.
I 3.
Within 30 calendar days of receiving a total of three violations of the UFSAR
[ Updated Final Safety Analysis Report) from the [NRC] during any [3-year) period, irrespective of the violation level, the operating license of the facility shall be suspended for a period of not less than 90 days and not more than 180 days.
4.
Within 30 calender days of receiving any harassment, intimidation and discrimination ("Hi&D") finding by the U.S. Nuclear Regulatory Commission, the U.S.
Department of Lsbor, or any [S] tate or [F]ederal court of competent jurisdiction, the operating license of the facility shall be suspended for a period of not I,.ss than 90 days and not more than 180 days.
i 9902270029 990211 PDR ADOCK 05000213 Q
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2-5.
If, within [5] years of a license suspensinn based on paragraphs 1 through 4 above, the licensee receives a total of three license violations from the [NRC), irrespective of the violation level; receives a total of three violations of 10 C.F.R. Par 150, including all applicable appendices, from the [NRC), irrespective of violation level; receives a total of three violations of the UFSAR from the.[NRC), irrespective of violation level; or receives any Hl&D finding by the [NRC), the U.S. Department of Labor, or any [S) tate or [Fjederal court of competent jurisdiction, the operating license of that facility shall be permanently revoked within go calendar days.
6.
In the event that the license of a facility is revoked pursuant to paragraph 5, no operation of that facility for the purpose of generating electric power shall be permitted during the pendency of any administrative orjudicial processes or appeals related to such revocation.
7.
In the event that the license of a facility is suspended or revoked under paragraphs [1]
through [5), the [NRC) shall designate an appropriate licensee to maintain the facility in shotdown mode for the duration of the suspension or until such tirne as a new licensee is found to operate the facility. [ Footnote omitted) NU [ Northeast Utilities) shall be responsible for all expenses related to the operation of the facility during such shutdown.
NU shall be required to post a bond in the amount of $500,000,000 ([5] hundred million) as reasonable assurance that it can fulfill this requirement.
The Pettioner further requested that these license conditions be imposed on the
' operating licenses of Millstone Units 1,2, end 3 before Commission approval to restart any of those plants, and further requested that these license conditions be imposed on the operating i
license of Haddam Neck before any decommissioning of that plant.
The Drector of the Off;ce of Nuclear Reactor Regulation has denied the Petition. The reasons for this denial are explained in the " Director's Decision Pursuant to 10 CFR 2.206" (DD 01), t!.s complete text of which follows this notice and is available for public inspection at the Commission's Public Document Room, the Gelman Building,2120 L Street, NW.,
Washington, DC, at the local public document rooms located at the Leaming Resources Center, Three Rivers Community-Technical College, New London Tumpike, Norwich, Connecticut, and at the Waterford Wry, ATTN: Vince Juliano,49 Rope Ferry Road, Waterford, Connecticut, for
- Millstone Units 1,2, and 3; and at the Russell Library,123 Broad Street, Middletown Connecticut, for the Haddam Neck Plant.
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A copy of the director's decision will be filed with the Secretary of the Commission for the Commission's review in accordance with 10 CFR 2.206(c) of the Commission's regulations. As provided for by this regulation, the decision will constitute the final actio'n of the Commission 25 days after the date of issuance unless the Commission, on its own motion, institutes a review of the decision in that time, Dete' at Rockville, Maryland, this 11th day of February 1998.
FOR THE NUCLEAR REGULATORY COMMISSION ue or Office of Nuclear Reactor Regulation i
-e EDO Principal Correspondence Control FROM:
DUE: 04/01/97 EDO CONTROL: G970140 DOC DT: 03/03/97 FINAL REPLY:
Ern00t C. Hadley
- TO:
Executive Director FOR SIGNATURE OF :
- GRN CRC NO:
Office Director DESC:
ROUTING:
REQUEST FOR LICENSING ACTIONS 10 CFR 2.206 Callan NORTHEAST UTILITIES (MILLSTONE UNITS 1, 2, AND 3 Jorden AND CONNECTICUT YANKEE) (FILED ON BEHALF OF ALBERT Thompson A. CIZEK)
Norry Blaha Cyr, OGC DATE: 03/03/97 Goldberg, OGC JKennedy, NRR ASSIGNED TO:
CONTACT:
Lieberman, OE NRR Collins SPECIAL-INSTRUCTIONS OR REMARKS:
NRR RECEIVED:
PARCH 4, 1997
~~~
NRR ACTION:
SP0: TRAVERS ACTION NRP ROUTING:
COLLINS vne DUE TO RRR D! REC.[0R'S Urru MIRAGLIA THADANI biY?I??^"
By 0asulLhl?l 1
J. KENNEDY
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- 040 0 MAiNSTRfEf PO t.OXbd9 wt St WARCHAM. M A 02% 76 t nNEST C H ADLty (508} 29ef 35.e or COUNSCL VOttf N PAf tuff 0LLEN F As (500) 296 700g WAITE P STUHL March 3, 1997 Leonard J.
Callan Executive Director of Operations U.S. Nuclear Regulatory Commission Washington, D.C.
20555-0001 re: Request for Licensing Actions 10 C.F.R.
% 2.206 Northeast Utilities Millstone Units 1, 2 and 3, and Connecticut Yankee
Dear Mr. Callan:
Pursuant to 10 C.F.R.
% 2.206, this letter constitutes a petition filed on behalf of Albert A. Cizek, formerly a Senior Engineer and Engineering Supervisor at Northeast Utilities
("NU"),
currently employed in NU's amployee concerns Program ("ECP"), to institute a proceeding under 10 C.F.R. 4 2.202, to modify the licenses issued to NU to operate the Millstone Nuclear Reactors, Unit 1, Unit 2 and Unit 3,
and Connecticut vankee by placing certain conditions, specified herein, on the operating licenses of each of those facilities.
As grounds for this petition, Mr. Cizek maintains that NU has knowingly, willingly and recklessly operated Millstone Unit 1,
Unit 2,
Unit 3 at Waterford, CT, and its Connecticut Yankee Nuclear Power Plant at Haddam Neck, CT, in violation of their respective operating licenses, the regulations of the Nuclear Regulatory Commission ("NRC"), and their respective Updated Final Safety Analysis Reports ("UFSAR") for a prolonged period of time, which unnecessarily but significantly compromised public health and saf ety by eroding the required def ense in depth philosophy; that NU has. knowingly., willingly and intentionally harassed, intimidatod and discriminated against its employees who. raise safety concerns in' violation of United States statutes and NRC regulations for a prolonged period of
- time, which also has unnecessarily but significantly compromised public health and safety by eroding the required defense in depth philosophy; and, that in the absence of express license conditions, there is no reasonable assurance cnat NU will cease and desist-n engaging in these activities in the
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The basis for the petition is set forth in further detail below.
Petitioner also requests that public hearings on the petition be scheduled in the immediate vicinity of the Millstone Nuclear Power Station and Connecticut Yankee Nuclear Power Station for the presentation of fuither evidence in support of the petition.
Petitioner specifically requests that these public hearings be held and decision on this petition icsued prior to the restart or decommissioning of any of these units.
Er999BAsUAc3nita..S9ndk119RS Petitioner proposes the operating licenses of Millstone Units 1, 2 and 3 and Connecticut Yankeel each be modified to i a.lude the following provisions:
1.
Within 30 calendar days of receiving a tota: of three license violations from the U.S. Nuclear Regulatory Commission during any three year period, irrespective of the violation
- level, the operating license of the f acility shall be suspended for a period of not less than 90 days and not more than 180 days.
2.
Within 30 calendar days ut receiving a total of three violat. ions of 10 C.F.R. Part 50, including all applicable appendices, from the U.S.
Nuclear Regulatory Commission during any three year period, irrespective of the violation
- level, the operating license of the f acility shall be suspended for a pe) iod of not less than 90 days.nd not more than 180 days.
3.
With.tn 30 cctlendar days of recei ing a total of three violations of the UFSAR from the U.S.
Nuclear Regulatory Commission during any three year period, irrespective of the violation
- level, the operating license of the f acility shall be suspended for a pe:ciod of not less than 90 days and not more than 180 days.
4.
Within 30 calendar days of receiving any harassment, intimidation and discrimination ("HI&D") finding by the U.S. Nuclear Regulatory Commission, the U.S.
Department of Labor, or any state or federal court of competent jurisdiction, the operating license of the f acility shall be suspended for a period of not less than 90 days and not more than 180 days.
1 / Although NU has made an economic decision not to restart Connecticut Yankee, licensed activities still are being conducted at that facility.
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- 5. If, within five years of a license suupension based on
_ paragraphs 1 through 4 above, the licensen receives a total of three license violations from the U.S.
Nuclear Regulatory Commission, irrespective of the violation level; receives a total of three violations of 10 C.F.R, Part 59, including all applicable appendices, from the U.S.
Nuclear. Regulatory Commission, irrespective of violation level; receives a total of three violationc of the UFSAR from the U.S.
Nuclear Regulatory Commission, irrespective of violation leve]; or receives any HIGD finding by the U.S.
Nuclear Regulatory Commission, the U.S.
Department of Labor, or any state or federal court of competent juris diction, the operating license of that f acility shall be permanentiv revoked wi*hin 90 calendar i
days.
- 6. In the event t. hat the license of a f acility is revoked pursuant to paragraph 5, no operation of that facility for the purpose of generating electric power shall be permitted during the pendency of any administrative or judicial processes or appeals related to such revocation.
7.
In the avent that the license of a facility is auspended or revoked under paragraphs one through five, the U.S. Nuclear Regulatory Commission shall designate an appropriate licensee to maintain the f acility in shutdown mode for the duration of the suspennion or until.such time as _ a new licensee is found to operate the tacility."
NU shall be responsible for all expenses related to the operation of the facility during such chutdown.
NU shall ne rcquired to pont a bond in the amount of
$500,000,000 (five hundred miillon) as reasonable assurance that it can fulfill this I
requirenent, Petitioner f urther requests that these conditions be imposed on the operating licenses of Millstone Units 1,
?
and 3
prior to Commicsian approval to start-up any of these plant.,, and further requests that '.hese conditions be impoud on the cpare. ting license of Connecti<.nt rankee prior to any deccmmission.ing of that plant.
2/ Since aver maintaining a facility in shedon mode is a licensed activity, it will be necessaty for the Comu!enion te designate a licensee to maintain the facility durirc nny such i
shutdown.
The only other alternative is for the NRC to cperate the facility-an alternative which is not-det.i rable given the commiucion's lax regulatory posture and which is of questionable legality.
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Imtionale_lon.. conditions Since August 1995 a series of public revelations about past operations of Millstone Units 1,
2 and 3 and Connecticut Yankee have eroded the public confidence in Northeast Utilities to safely and legally operate any or all of its nuclear power plants.
Corresponding revelations about the abdication of regulatory responsibility by the NRC have similarly eroded the public confidence in the NRC to adequately protect the public health and safety.
[
The only proper conclusion to be drawn from these revelations is that NU is unfit to hold any license to operate a nuclear power plant.
Toward that
- end, one group of Petitioners-Citizens Aworenes s Network ( "CAN" ) and the Nuclear Information and Resource Service ("NIRS")-has filed a petition under 10 C.F.R.
% 2.206 to revoke the operating licenses of si) HU plants in Connecticut.
- However, reality dictates that the petition has no chance of
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success.
- Indeed, the NRC has yet to rule on the requested licensing action in a petition filed by George J.
Galatis and We the People, Inc., of the United States in August 1995, despite the fact that every allegation in the petition has been proved true, has been verlfied by the NRC Office of Investigations or Office of Incpector General, and currently is the subject of a criminal probe by the United States At torney for the District of Connecticut.
i Since there is virtually no chance the NRC will take the k
appropriate licensing action, the Petitioner requests that, in lieu of such action, the Commission impose certain self-executing conditions on the operating license if each facility.
From a nracticai perspective, the NRC has given the public only an illusory process in 30 C,F.R.
% 2.206.
Two of the permissible requested actions -revocation and suspension of a license-cannot be
(
granted.
Even to maintain a nuclear power plant in a shutdown mode requires the performance of licensed activities.
The resulting real.ity is that the NRC cannot revoke or suspend an operating license becauce the icensee ther could be unable to pertorm the lironsed activities required to ef ely mainta b. che shutdown plant.
Thus, an operating license can be revoked only if the NRC finds another qualifica licensee to oversee operations during shutdown.
q That is a highly unlikely event given the f act that shutdown plants generate no revenues but conversely genernte extremely high
- expenses, h.
Also, because t tle NRC has a] N ed c.n e utility to construct and operate four nuclea! rower plants in cuch a small geegraphic area, re'!ocation M UUd n licenses would likely cauce economic chaos in the State of Connecticut.
Loss of jobs, the cost of replacement power and the ultimate cost of deccmmissioning would end up
'cenalizing the residents of Connecticut, who have innocently been g
exposed to unnecessary health and safety risks, instead of NU and
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itu rMfrcers cnd executives, who have engaged in deliberate and i
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J willful misconduct.
NU has promised, time and time again, to address its failings without success.
Although in different positions, most of the individuals responsible for these failings remain as part of NU management or staff.
Executive NU management maintains its position that the past should be ignored in restart considerations.
This position apparently is embraced by the NRC as evidenced by its willingness to commit virtually all of its resources to working with NU on its restart plans, and virtually none of its resources to punishing NU for itc past transgressions.
Moreover, NU is attempting its recovery with primary reliance on loaned, short-term management that may not be around to provide the required checks and balances once normal operations resume and in the long term.
Common sense and logic dictate that there is no reasonable assurance that NU will cease and desist from violattug its operating licenses, the regulations of the NRC and its UFSAR's.
Nor will it cease and desist from harassing, intimidating and discriminating against its employees who raise safety concerns.
i Trust han been lost.
Only a sustained period of operating within the parameters of the law, NR^ regulations, license and UFSAR requirements, without deviation or set back, can restore that trust.
For these reasons, and because the regulator cannot be trusted to exercise good judgment on behalf of the public it is mandated to protect, the placement of precise limiting conditions, along with adequate ' financial precautions, on the operating licenses of Millstone Units 1,
2 and 3 and Connecticut Yankee is the only logical means of attempting tc ensure the plants are operated safely and legally in the future.
Thus, aJthough the Petitioner firmly believes that NU has demonstrated that it is unfit to hold any operating licenses, Petitioner instead seeks modification of the licenses of Millstone Units 1,
2 and 3 and Connecticut Yankee as the only logical safeguard.3 Cont 3ntioDs of PeAition Petitioner seeks the limiting conditions on the operating licenses of Millstone Units 1, 2 and 3 and Connecticut Yankee based on the following contentions:
O Contention No.
1: NU has knowingly, willingly and recklessly operated Millstone Unit 1, Unit 2, Unit 3 at 3 / Petitioner does acknowledge that the ultimate ef fectiveness of the license conditions depends heavily upon proper regulation by the NRC.
There is, in Petitioner's opinion, no concrete evidence that such regulation is forthcoming.
However, the NRC's f ailure to adequately regulat.e io not properly the subject of a petition under 10 C.F.R.
ti 2.206.
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Waterford, CT, and its Connecticut Yankee Nuclear Power Plant at Haddam
- Neck, CT, in violation of their respective operating licenses, the regulations of the NRC, and their respective UFSAR's for a prolonged period of
- time, which unnecessarily but significantly compromised public health and safety by eroding the required defense in depth philosophy.
O Contention No.
2: NU has knowingly, willingly and intentionally harassed, intimidated and discriminated against its employees who raise safety concerns in violation of United States statutes and NRC regulations for a prolonged period of time, which unnecessarily but significantly compromised public health and safety by eroding the required defense in depth philosopny.
1 o Contention No. 3: In the absence of express license conditions, there is no reasonable assurance that NU will cease and desist f rom engaging in these activities in the future.
Each of the contentions is addretsed below.
Contstnt. ion _No_l The evidence is overwhelming that, for several years now, NU has operated as a " rogue utility" deliberately placing cost control over safety and intentionally violating the 6 ditions of its operating licenses, NRC regulations and UFSAR'u.
For example, under It C.F.R.
S 50.73, NU, like all licensees, is required to tain specified events in License Event Reports ("LER").
report c Most of the events required to be reported by the regulation specifically relate to license violations, regulatory violations, UFSAR violations, e.g.,
operations outside of a plant's design
- basis, or violations of the principles of sound engineering judgment.
By its own admissions through LER's, which have been incomplete, inaccurate, misleading and intentior. ally false, NU has an abysmal record.
Over the last four years, the following number LER's have been filed for Millstone Units 1,
2 and 3
and Connecticut Yankee:
LEAR WIT WIT WIT UNIT MP1 MP2 MP3 CY 1993 25 23 23 19 1994 33 43 15 29 1995 32 45 22 23 1996 65 41 50 30 6
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The industry average for LER's per plant for 1993 through 1995 has been 13, 12 and 11, respectively.
See, Annual Report, 19 9 4-FY 9 5, Of fice for Analysis and Evaluation of Operational Data (July 1996) at p.
40.4
- Moreover, NU, at all of its Connecticut plants, has violated license and regulatory requirements, cast aside good engineering judgment and intentionally operated the plants outside of their design bases.
Revelations of this type of conduct in the past 18 months are too numerous to fully recount in this petition.
However, because extensive documentation already exists of these examples, Petitioner requests that the NRC take administrative notice of certain documents listed in the section
- entitled,
" Evidence in Support of Petition," infra.
But Petitioner's raquest for modification of the operating licenses of Millstone Units 1, 2 and 3 and Connecticut Yankee also is based on personal experience.
In one matter in which Petitioner was directly involved as the alleger, NU willfully operated Millstone Unit 1 for over 20 years without testing, as required, certain containment isolation valves, including CU-29.
Petitioner originally challenged the decision of NU to place testing of CU-29 in the Integrated Safety Assessment Program
("ISAP").
ISAP, as Petitioner pointed out to NU, was designed to deal with new and emerging regulatory issues that had arisen since the licensing of the plant.
Testing of CU-29 was a requirement by virtue of the NRC's adoption of 10 CFR Part 50, Appendix J in Fe'iruary 197 3.
Furthermore, it became a license condition when the f ull term operating license ("FTOL") was granted in 1986.
Thus, it was clearly part of the licensf ag basis of Millstone Unit 1 and, as such, was improper for inclusion in ISAP.
Further, Petitioner challenged the designation of testing of CU-29 as a " low priority" in ISAP.
When, because of Petitioner's continued insistence, the valve was tested in 1995 it was determined that the valve leaked excessively and could not perform its intended saf ety f unction.
Significantly, in a report by the U.S.
Nuclear Regulatory Commission Office of Inspector General ("OIG"), entitled "NRC STAFF ACTIONS TO ADDRESS CU-29 ISOLATION VALVE ISSUE," Case No.96-06S, dated September 3, 1996, the OIG stated:
The OIG learned that in February 1973, the NRC issued Appendix J " Primary Reactor Containment Leakage Testing For Water-Cooled Power Reactors",
which became a
requirement for all licensed nuclear power reactors.
Specifically, Appendix J requires that all operating licensees for water-cooled power reactors test the leak-4 / No data is currently available for 1996.
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tight' integrity of. the primary reactor containment, systems and components, including containment isolation 4
valves.
Appendix J-tests are required to be performed.
-during each reactor shutdown for refueling but in no case at intervals. greater than two years.
.In August 1975, NRC requected NU to determine whether Millstone Unit I was in full compliance with Appendix J and if not, to identify planned actions and to prepare a schedule to cchieve compliance.
NRC ndvised NU that possible courses of action included modifications to design features to permit conformance with the testing requirements as well as requests for exemptions from Appendix J requirements.
In November 1975, in response to the NRC's
- request, NU provided a
summary of containment isolation valves and identified tests conducted to that point.
NU also identified valves which-would require exemption f rom Appendix J requirements.
Between 1975 and 1984, the NRC staff and NU exchanged correspondence regarding the status of NU's compliance with Appendix J.
i In November 1984, NRC initiated the Integrated Safety Assessment Program (ISAP) to conduct integrated assessments for operating nuclear power reactors.
The ISAP was intended to address plant-specific evaluations of licensing actions, plant improvements and unresolved generic safety issues.
Millstone Unit I was one of two operating plants selected by the NRC to participate in the ISAP pilot program.
[I]n April 1988, NU requested exemptions relating to Appendix J
for certain containment penetrations, including containment isolation check valve CU-29.
NU U
requested an exemption from testing requirements for check valve CU-29 because design features of the check valve did not permit testing unless certain modifications were made.6 In June 1991, the NRC denied this exemption request.
Millstone Unit I was shut down for refueling outage 13 when NRC denied the exemption reque-t; however, at the end of outage 13 the licensee reau.aed plant operations without testing valve CU-29, Between October 1992. and June 1995, NRC reviewed and 1
concurred on several of NU's ISAP reports which outlined plans to modify and test valve CU-29 in accordance with Appendix J during refueling outage 15.
Ultimately, when 5 / What the Inspector General Report does not make clear is that NU twice requested exemptions for CU-29 and was twice turned down by the NRC.
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. O tested during refueling outage 15, NU determined that valve CU-29 leaked excessively and may not have been capable of performing its intended containment isolation function.
Id. at pp.
4-5.
Even more significant is the fact that CU-29 was designed to provide containment isolation in the event of a single active failure to CU-28, a motor-operated valve.
It has been determined that CU-28 was "not environmentally qualified to perform its containment isolation under adverse conditions. "
Id. at p. 16.
As such, this penetration and primary containment were outside of design basic, which compromised public health and safety by eroding the defense in depth philosophy.
Likewise, Unit 2, has seven containment isolation valves which lack environmental qualification. See, Combined NRC Inspection Report Nos. 50-245/96-06, 50-336/96-06, 50-423/96-06.
In a letter to Ted C. Feigenbaum, Executive Vice President and Chief Nuclear Officer, NU, dated October 9,
1996, accompanying the inspection report, Wayne D.
Lanning, Director, Millstone Oversight Team wrote:
The third apparent violation at Unit 2 involved the lack of environmental qualification for seven containment isolation valves that must be re-opened during the post-accident phase of an accident.
This issue is of particular concern because qualifications of four of the seven valves were the subject of specific NRC review in 1983, and the safety function for these valves was not identified and corrected at that time.
Further, the NRC identified other weaknesses in the implementation of the environmental qualification program that raise questions regarding the ability of EEQ components to perform their safety f unction. Therefore, NRC consid.ers the completion of outstanding EEQ and high energy.ine break program activities, and the revalidation of rae qualification of affected components to be a plant start-up issue.
Id.
at p.
2.
Thus, as with past infractions, problems which surface at one unit usually are indicative of problems that exist 6 / In an LER filed on April 25, 1996, NU repo rt.ed that at Millstone Unit 2 the safety functional requirements ("SFR's") of the seven Electrical Equipment Qualification
("EEQ")
solenoid operated valves ("SOv's") could not be demonstrated.
The report noted the "SFR's indicate that these SOV's are required for containment iaolatic" and subsequent post-accident operation. " LER 96-019-00, April 25, 1996 at p.
1.
The incident is safety significant because NU "cannot demonstrate reasonable assurance that the 7 listed EEQ valves would have performed their safety function on demand post-accident."
Id. at p.
4.
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at NU's other ur.its.
In addition to evidence that problems that are identified at one unit also exist at-other units, there is overwhelming evidence thct
- multiple, related problems have been allcwed to exist that compromise safety beyond any acceptable limito.
For example, in conjunction with the problems previously outlined with regard to CU-29 and CU-28, piping system welds associated with those valves suffered from intergranular stress corrosion cracking ("IGSCC")-a phenomenon of which the industry has been aware for many years and which was the subject of NRC Generic Letter ("GL") 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping."
The above-cited inspection report concluded:
During the RFO 15 inspection, thirty five (35) welds were examined containing IGSCC. Fourteen of the 35 welds were 7
previously identified by the UT Level II technician, as early as 1984, as having at least one IGSCC indication, 8
and were subsequently overturned by the NDE Level III.
The disposition of the 35 welds with IGSCC, during RFO 15, was to replace the component or weld overlay the components prior to start-up.
Six reactor coolant components (RCAJ-2, RRJJ -4, RREJ-4, RRCJ-4 and CUBJ-18) with flaws were placed inservice, between 1984 and 1995, without flaw analysis as required by ASME Section XI, 1986 Edition, Paragraph IWB-3640.
During the U"' examinations, each component had at least one IGSCC indicction.
The ASME Section XI analysis was not performed on the com';onents because the UT Level III inappropt j ately evaluated the IGSCC indications to be geometry.
The indicat. tons were determined to be cracks during refueling outr.ge 15.
The licensee performed an evalcation during the November 1995 refueling outage, RFO 15, in accordance with ASME Section XI, 1986 Edition, IWB-3640, to determine the operability to the components.
The licensee determined the components did not meet the requirements for continued service and declared the components inoperable.
The licensee defined inoperability of a component as a decrease or elimination of the operating safety margin for structural integrity.
The licensee determined the safety margin is decreased when a crack through wall dimension in the component is 7 / " Ultrasonic Testing."
8/ "Non-Destructive Examination. "
It should be noted that the NDE Level III is an industry cenified person.
The number of
" errors" made by the NDE Level III in this case is strongly indicative that the " errors" were intentional.
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equal to or greater than 75% of the pipe wall nominal i
I thickness.
The six components had intergranular stress corrosion cracks (ICSCC) that were greater than 75% through the wall.
Two of the six components leaked during preparation for weld overlay.
e reactor coolant systems were degraded to the extent a detailed evaluation was necessary to determine system operability.
The results of the licensees (sic) evaluation determined the six components had an unacceptable structural integrity and a high probability of abnormal leakage.
Id. at % M1.2b, ISI ProgILm_ Review (emphasis added).
Of particular significance is the finding with respect to pipe weld CUBJ-18 because of its direct connection with CU-29.
In Licensee Event Report ("LER") 95-029-00, dated December 15, 1995, NU stated:
A structural review was performed for CUBJ-18 and concluded that operability may not have maintained in the event of a design basis seismic event.
In the worst case, and the weld had failed, leakage [ rom the reactor would have been limited by check valve 1-CU-29.
This valve was inspected during this outage and verificd that it would '1 ave shut and prevented gross leakage from the penetration.
Id. at p.
4 (emphasis added).
In other words, one of the pipe welds most susceptible in the event of a seismic event was in the worst possible location, i.e., between CU-29 and CU-28, and outside primary containment.
Although NU, in LER 95-029-00, took credit for an operable CU-29 check valve, in LER 96-012-00, dated March 7, 1996, NU conceded:
On December 3,
1995, with the plant shutdown and the reactor in COLD SHUTDOWN condition, it was determined that a reactor water cleanup system containment isolation check valve, 1-CU-29, had exceeded its maximum rate leak while it was in operation.
Id. at p.
1 (emphasis added).9 NU concluded that there "were no adverse safety consequences" as a result of CU-29 being inoperable "since the redundant valve in the penetration was capable of performing the containment isolation function."
Id. at p.
2.
In making this determination, NU deliberately and intentionally omitted any consideration of the IGSCC that had occurred to pipe 9 / Nowhere in LER 96-012-00 does NU mention its earlier findings regarding IGSCC and pipe weld CUBJ-18.
11 i
l
e i
weld CUBJ-18 per LER M-029-00, despite the fact that the design basis for primary containment includes a " design basis earthquake (DHE)" according to the UFSAR.
See, UFSAR at Sec.
6.2.1.1.1.
Not only do the above-described events constitute a course of deliberate and intentional micconduct, ac well as deliberate and intentional material falso sta!ements, but it is clear that NU, through these actions, significantly jeopardized the public health and safe *.y.
Far irom the "defonce in depth" strategy employed in the design and operation of nuclear power plants, NU allowed Millstone Unit 1 to operate in conditions whete a single failure would have resulted in an uncontrolled release of radiation.
Pipe weld CUBJ-10 had en excessively deep crack which made the piping rystem inoperable; simply stated, an earthquake could result in a failure of the piping system allowing a direct release path of reactor coolant or containment atmosphere autoide primary contair. ment through leaking CU-29 with no means to isolate.
Petitioner also han direct knowledge that the deliberate and intentional disregard for license requirements, NRC regulations, good engineeri a.
judgment and the design basis of the Millstone
'inits continue-in the present under NU's new management.
For example, during the November 1995 refueling outage, Petitioner obnerved " scale" in the Low Pressure Coolant Injection ("LPCI")
heat exchanger tubing after the tubing was hydrolazed.
Problems with scaling had been identified and addressed by the NRC in GL 89-13,
" Service Water System Problems Affecting Safety Related Equipment," dated July 18, 1989, and GL 89-13, Supplement 1,
dated April 4,
1990.
According to Northeast Utilities Incependent Root Cause Evaluation, A/R 96008360,
"!'P1 LPCI Heat Exchanger Tube Scaling," appro ud July 11, 1996:
During 1989/1990, the NRC issued GL 89-13 and Supplement 1.
The GL identified tubing scale (e.g.,
calcium carbonate depocits) as one type of fouling known to degrade service water heat exchanger performance.
The plan outlined in GL 89-13 intended that the industry perform thermal performance testing of their safety related service water heat exchangers The general response by the industry to GL 89-13, including NU, was to delay its implementation and/or pursue low cost alternatives to the thermal performance testing plan outlined by the NRC.
lU While NU u t.s takjig FRM
- approach, the NRC was conducting Service Water System Operational Performance inspections (SWSOPI's) throughout the nuclear power industry that included an assessment of utility responses 10 / " Frequent Regular Maintenance."
12 l
_ _ _ ~ _ _ _ _ _ _ _ _ _
to GL 89-13.
Early findings from these SWSOPI's were reported to the industry and highlighted to MP and CY j
Engineering Groups in memo ES-ME-94-076, dated February i
l 18, 1994. This memo pointed out that the FRM approach was being seriously challenged by the NRC, basically because the NRC was finding an absence of a comprehensive, documented evaluation of each heat exchanger's capability to meet its design thermal requirements.
Id. at pp.
5-6.
During the November 1995
- outage, Petitioner questioned the 3
potential nafety impact of the scaling on the heat exchangers, e
partjeularly given the low safety margin of the LPCl heat would be conducted prior to the end of the outage.gmance testing exchangers.
Petitioner waq told that thermal perf However, NU deliberately and intentionally took no action to perform such tes**ing and, instead, p.epared to restart without conducting the test.
.At that time (January 1996),
Petitioner discussed the cituation with the NU Employee Concerns Program ("ECP") which confirmed the low safety margin.
However, ECP took no further 4
action.
4 j
Finally, on March 26 1996, Petitioner initiated Adverse Condition i
Report ("ACR") 9002 declaring the LPCI heat exchangers inoperable, The Root Cause Analysis referenced above also found that Millstone j
Unita 2
and Connecticut Yankee had no definitive inspection procedures for heat exchangers and that Millotone Unit 3 only had t
a visual inspection program to detect macrofouling by b;le mussels.
Ed. ht p.
6.
When it became apparent that NU still would not conduct the required toot, Petitioner brought his concerns to the NRC, Region 1.
On May 14, 1996, Richard Cooper, Director, Division of Reactor Projects, Region 1, referred Petitioner's allegation regarding LPCI back to NU through Ted C.
Folgenbaum, Executive Vice President, Nu clea r.12 Although Mr.
Teigenbaum was advised that the distribution of the allegation should be controlled and " limited to personnel with a
'need to know,'"
the allegation received I
i.
11 / It is.important to note that, at this time, Millstone Unit j
1 was still in a planned refuel outage, That outage later was extended by a 10 C.F.R.
% 50.54(f) letter and placement of Unit 1 on the NR ' " Watch List."
The plant was later degraded tn a Category 3 plant and may not restart until given approval by a vote of the Commission.
12 / The referral was contrary to the NRC Volume 8, Licensee 0"ernight Programs, Management of Allegations Handbook.
The NRC's referral of the allegation back to NU currently is the subject of an investigation by the Office of Inspector General.
13 h'
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i widespread distribution at the Millstone site.
This fact was acknowledged in the Independent Root Cause Evaluation cited above, which stated "[bjecause of the uncontrolled distribution of the NRC transmittai,-personnel involved with the LPCI scale issue appeared guarded and, at times, focused more on the NRC allegation than on the scale issue and any generic implications. "
1d. at Addendum, p.
j 1.
One of the personnel involved with the Independent Root Cause Evalt.a tion specifically referred to the matter as the "Cizek Affair."
The same Root cause Evaluation recommends that
" future NRC j
allegation issues be evaluated by a third party until NU.
. has the forthrightness to handle such issues appropriately."
Id. at Addendum, p. 1.
After continued persistence, Petitioner convinced i
the ECP Director to retr.rn " referred allegations" back to the NRC.
!!owever, the director's decision was overturned by Mr. Feigenbaum.
4 Subsequently, at Mr. Feigenbaum's direction, " referred allegations" j,
are addressed as standard practice within Nuclear oversig' t without the involvement of ECP.
1 According to a
subu quent root cause analysis performed by Millstone Jnit Engineering, "the reason for not assessing the affect oa heat exchanger performanco is the lack of having compreh<insive heat exchanger
- testing, monitoring, and visual inspection programs with clearly defined acceptance criteria "
j See, "Nertheast Utilities Root Cause Investigation: Heac Exchanger Tube Sice Scale Formation, " Rev.
2, at p.
2, dated December 12, j
1996.
Typical of NU's intentional disregard of safety issues is the conclusion that the root cause was the failure to have an effective program.
The root cause analysis fails to ask the obvious question: Why wasn't there such an effective program?
The 4
answer:
Because NU had deliberately delayed developing and implementing such a program even after being advised by the NRC i
that the FRM approach was questionable, if not completely inadequate, i
i I
Other examples of NU'r deliberate misconduct abound.
For example, l
in Notice of Violation (NRC Combined Inspection Report No. 50-l 245/95-31; 50-336/95-31; 50-423/95-31), dated December 7,1995, the Commission decided to excuse NU's conduct for two significant violations at Millstone Unit 1.
The first violation involved an existing single failure vulnerability in the loss of normal powcr logic that would have prevented both emergency power sources from properly starting and sequencing required loads.
This constitutes a violation of your technical specification j
- L
'we in the event of a single failure, following a l
loss of normal power during a Loss of Coolant Accident l
(LOCA), a loss of both emergency power sources would occur.
The violation was caused by your inadequate l
design reviews of two different modifications installed j-14 I
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I in 1976 and 1989, which failed to detect the single-
}
fallure vulnerability.
j i
l The second violation involved two examples of existing vulnerabilities in the standby gas trea tinent e
j system (SGTS).
In the first example, the system could rupture if a LOCA occurred while venti"q the drywell, which would result in a complete loss of.iJS because the c
drywell isolation valves would not cine in tirne to L
prevent the pressure wase generated from the LOCA from affecting the integrity of both trains of the system's filterc*
- Second, an existing singic. failure vulnerability in the SGTS would prevent the systom irom mitigating the effects of a LOCA.
Specifically, during a LOCA, the inability to isolate one train on the SGTS, l
if it were to fall in conjunction with the inadequate backdraft damper design, would allow a short cycle flow path to be established.
This flowpath would prevent establishing or maintaining the reactor building negative
= pressure, which is required to prev ( ~ a post LOCA ground level release.
Id. at pp.1-2 (emphnis u ded).
The violet. ions regarding the SGTS are particularly significant since that is one of the major systen.s that would have been relied upon in the event of a release outside i
primary containment due to the previously cited CU-29 scenario.
In short, NU operated Millstone Unit 1 in such a way that fundamental design basis assumptions could result in an unacceptable post-LOCA i
ground level release.
Simply stated, multiple systems were outside theit. design basis which reduced safety margin and compromised i
public health and safety.
Fortunately, no design basis accident occurred.
[-
In an inspection between November 5 and December 26, 1995, NRC resident inspectors found other violations that are indicative of l
~ 1icense requirements.
For example, at Millstone - Unit 1,
during Intentional misconduct, or at the least reckless disregard for shutdown:.
The operations staff f ailed to prevent work that had the potential for draining the reactor vessel while fuel removal-was in
- progress, as required by technical-specifications.
The licensee does-not-have an adequate c
- process to ensure all applicable technical specifications are. implemented during refueling. - A violation was cited
- because it is an example of continuing problems with inadequate implementation of current regulatory, i
requirements, The-licensee's subsequent position that 15 y%m...u-9-w--.py-sa-.
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13 establinhed a
50 GPM leak as threshold for the potential for draining the reactor vessel was inconsistent with the licensing basis that establishec i
the applicable technical specification, l
3 operator actiors to raise reactor pressure above the initial pressure assumed in accident analysis reflected a lack of understanding and respect for the plant's design basis.
[T]his issue is significant because 1
operatora purposefully changed the normal operating i
pract J ce wtthout implernen ting appropriate procedural controls including the necessary safety evaluation. In
- addition, the licensee failed to implement prompt procedure changes when the lack of procedural control over operat.ing reactor pressure was identified on two prior occasions.
Executive
- Summary, Millstone Nuclear Power
- Station, Combined I n s})ection 245/95-42, 336/95-42, 423/95-42, February 6, 1996 at p.
11 (emphasis added).
The same inspection report found that, during the November 1995 shutdown of Unit 3 "[a]n unplanned dilution of the reactor coolant system occurred due to inadequate procedures, poor communicetions between the chemistry and operations Department, and poor maintenance of the letdown high temperature divert valve."
The report noted that this was the fourth unplanned dilution event at Unit 3 since December of 1993.
Jd. at p. iii.
In two other instances at Unit 3, the inspectors found:
At Unit 3,
licensee management displayed a lack of conservative safety perspective in not validating the conditions that exited with a leaking 10-inch unisolable check valve in RCS.
The associated operability determination was based on engineering judgment and assumptions, and was not entirely accurate.
[T]he socket weld f ailures should have been prevented by corrective actions for similar vibration induced f ailures identified in 1994.
The licensee postponed these socket weld repairs during the 1995 refueling outage, based in the inappropriate safety judgment that any socket weld failures would only result in a small RCS leak.
Id. at p v, A December 13, 1995 LER provides another example of NU deliberately deciding to tolerate conditions that result in licence violations.
13 / " Gallons per ninute."
16
4.
Je f.*
1 Thin time, the event involved Millstone Unit 2 where reactor core power exceed the rated thermal power of 2,700 megawatts for a period of some 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />.
According to the uER:
i j
On November 15,
- 1995, at 1045 hours0.0121 days <br />0.29 hours <br />0.00173 weeks <br />3.976225e-4 months <br /> with the - plant -
j operating in 14 ode 1 at 100% power, a review of plat i
operating parameters identified that the reactor core l
thermal power level had inadvertently exceeded the maximum power level permitted by the operating license.
l This event was csused by incorrect steam generator J
blowdown ilow rate value in the core heat balance i
calculation.
This caused the calculated core thermal j
power to be less than the actual core thermal power.
i
,c.R 95-043-00, December 13, 1995.
As part of its solution to the l
- problem, NU proposed simply that "the installation of process instrumentatio!.-to measure the blowdown rate ir being evaluated 4
i Jd.
The root cause of this event was a combination of l
inadequate design
- control, inadequate operator training and inferior, cost conscious corrective action-all of which are brought about by NU's intentional and deliberate disregard for the 1411\\ stone Unit 2 licence and UFSAR and NRC regulations.
Other safety significant issues involving he determination of actual core power occurred at !4111 stone Unit 1 and Connecticut a
l Yankee.
Both involve the measurement of feedwater flow which is a j
key input into the calorimetric calculation which determines actual core power.
In accordance with the operating licenses and i
technical specifications, actual core power must not exceed specified values to maintain the validity of assumed accident analyses described in the UFSAR's.
J Shortly after the initial start-up of 14113 stone Unit 1 in December
- 1970, problems with the measurement cf feedwater flow were i
encountered due to erratic and inconsistent differential pressure measurement.s from the primary flow elements-in this case Venturi tubes.
As a short term, but questionable, solution, turbine first stage
- pressure, in conjunction with its corresponding flow cotfficient and other parameters, was used to approximate steam flow.
a 4'
This is not the preferred method to treasure primary flow for use as an input into the calorimetric calculation for a variety of reasons."
These reasons drive the calorimetric calculation in i
M / 14ost important, knowledge of the moisture content of the steam is required.
This is a difficult parameter to measure, so the unverified design or calculated value was used, although this was subject to a large error.
Additionally, steam flow was used subject to unmeasured-or unknown losses upstream of the turbine i
17 i
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i a non-conservative, unquantified direction as compcnents age and degrade.
Regardless, a rigorous uncertainty analysis to bound core 1
l power error and assess the impact on assumed accident analyses described in the UFSAR was not performed.
The measurement was used well into the 1980's, when feedwater correction f actors as large as nix p;ercent were applied to " adjust" measured feedwater flow.
At th _ time, Petitioner, then an established engineering supervisor, and his immediate manager, discussed the subject and all of its inherent weaknesses with i
Millstone Unit 1 Engineering, but were rebuffed..
Not until Petitioner conveyed the results of turbine cycle heat balance calculations, which suggested a net core power and corresponding electrical generation increase, did Millstone Unit 1 opt to address i
the problem.
After replacement of the feedwater Venturls, core power decreased about one percent, meaning that 3ast operation was in excess of licensed power.
The change could have been much larger and more significant.
- Sometime prior to 1986, Connecticut Yankee encountered-problems with the measurement of feedwater flow oy its primary flow elements-in this case orificed flow sections.
Ultimately, one of the four orifice plates had to be replaced.
However, Connecticut Yankee chose to manufacture its own orifice plate and install it without calibrat. ion.
All Quality Assurance ("QA") requirements wero ignored.
Upon discovery by Petitioner's subordinates, Connecticut Yankee had no choice but to take action.
During a subsequent refueling outage, all four orificed flow sections were replaced.
Quite remarkably, comparison of the calibration results from an accredited flow laboratory showed only a small difference.
- However, the difference could have been quite large and significant.
Connecticut Yankee essentially ran " blind" for an-unknown period of time since a rigorous uncertainty analysis to bound core power error and assess the impact on assumed accident analyses described in the UFSAR could not be performed.
In both cases described above, NU demonstrated intentional and deliberate disregard for the Millstone Unit 1 and Connecticut Yankee licenses and UFSAR's, and-NRC regulations.
At Millstone Unit 1, this method was used in excess of 15 years.
At Connecticut Yankee, the duration is unknown.
f In an LER filed on Decemoer-13,1996, NU reported at Millstone Unit 3, the " plant had opsrated in a condition that was outside the design basis due to _a def J ciency in specific design conditims for
=
which could be significant, primarily due to turbine bypass valve leakage to the condenser.
A l s o,- the turbine first - stage flow coef ficient is ' subject to change as the first stage goometries enlarge. due to degradation,. prinarily due to erosion which is proportional to moisture content.
18 m
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a system needed to remove residual heat and mitigate the consequences of an accident.
It was determined that the containment Recirculation Systen. (RSS) spray piping and supports were not adequately designed for thermal loads resulting from accident ten 'ratures."
LER 96-007-02, December 13, 19 9 6 at p.
1.
This problem existed as part of the original plant design.
Id. at p.
3.
- l. gain, the condition was sign 2ficant "in that had the plant experienced a design oasis accident in containment such as a LOCA or IIELD, than the potentini existed that these systems may not have been able to f ulfill their required saf ety f unction. "
Id. at p. 4.
!!oreover, IJU had to acknowledge thot there were two similar events as reflected in LER 96-006-00,
" Plant Shutdown Required by Technical Specifications for Auxiliary Feedwater Containment Isolation Valves Declared Inoperable" and LER 94-006-00, " Auxiliary Feedwater Pipe Restraints, Inadequate Design Due to Design Error."
Id. at p.
5.
Likewise, both of these deficiencies were original design deficiencies.
Also, at Unit 3,
14U's intentional and deliberato decisions to ignore problems resulted in repeated leaks in four Reactor Coolant System ("RCS")
loops through vibrational fatigue of small bore piping.
This generally reiers to a smaller diameter pipe (branch piping) welded to a socket (receiver,) which is, in turn, welded unsupported to a larger diameter pipe (main piping).
Vibrational fatigue occurs when a stimulus, such as a running pump, vibrates the smaller diameter pipe and ultimately, through cyclic fatigue the smaller diameter pipe to socket weld will fail resulting in leakage.
At 11111 stone Unit 3,
a small diameter pipe connected to the four RCS loops is used to locally measure pressure at sixteen locations.
One weld failure resulting in leakage occurred May 1992.
Two indications (f aults) without leakage and one weld f ailure resulting in leakage occurred September 1994.
One weld failure resulting in leakage occurred December 1995.
In September 1993, vibration testing was conducted to determine failure susceptibility.
See, Memo from G.E. Dreschler, component Test Services, to M.D.
- liess, re:
" Millstone Unit 3 Reactor Coo ant Piping Vibration Testing Flow Instrumentation Lines."
The resiits showed negligible apparent susceptibility due to the lack of measured excessive vibration amplitudes.
The report coreluded that no further testing was necessary.
However, as noted, another failure occurred in September 1994.
Consequently, a
non-destructive examination was performed to identify faults as a precursor to failure.
Two faults were identified.
Unfortunately, existing non-destructive examination was not foolproof and all faults may not have been detected.
Consequently, further review was required.
Additional examination of the September 1994 f ailure by an outside 19
i j s a
i source, ABB Combustion Engineering Nuclear Operations, confirmed cyclic fatigue, see, " Examination of the Socket Weld Crack in i
Millstone Unit-3 Loop C Pressure Top V125,* Draft Report, MISC-PENG-TR-028, December 1994.
NU deleted two key points in tne final versio.."
First, the final repcet deleted a notation in the j
draft that indicated:
1 The US Nuclear Regulatory Commission (NRC) expressed their concerns to NUSCO about the cause of the failure.
l
{
Id. at p.
1-2.
But more important, the following conclut;1on was i
deleted from the final report:
i Decause the f ailure of one of these welds could result in a significant safety problem, it is recommended that each of the 16 similar wolds be visually inspected every outage for the presence of boric acid deposits.
Radiographing the welds has proven to bo difficult and 1
may not yield correct results.
A review of any maintenance performed in the vicinity of the elbows may i'
provide some insight into the cause of the failure.
It is recommended that the vibration of the lines be period of several days during normal monitored for a i
operation of the plant.
l Id. at p.
S-2.
Based on th" above evolving scenario which demonstrated continued l
vulnerability, the Welding and Materials Engineering Section "alrangly recommends that MP3 thoroughly review the existing piping design of the RCS instrument lines.
Piping design modifications to eliminate socket welds or additional support modifications may be necessary to eliminate fatiguo loading."
Memo from A.J.
- Silvia, Component Engineering Services to George Pittman, MP3 Engineering l
Director, March 17, 1995, "MP3 Reactor Coolant System - Socket Wold t
Failure (3RSC*V125)".
Instead, NU management defended a past operability determination indicating the problem was of low safety significance since identification of nocket weld failures is not unusual in the i
nuclear industry, socket weld failures are the most common weld failure, use of socket welds '. ave been restricted _to small lines, l
socket weld failures have demonstrated leak before
- fcilure, unidentified reactor coolant leakago is monitored by a system capable of detecting a 1 gpm change, there is -a Containment Atmosphere Gaseous and Particulate Radioactivity Monitoring System i
to monitor RCS leakage, any leakage would be collected within the containment drain system, the_3/4 inch diameter instrument line is limited by a 3/8 inch diameter port size, and normal makeup has i
"/ The final version of the report is dated February 1995.
l 20 l
L'
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.j i
sufficient capacity to maintain preasurizer level and compensate for the problem.
- See, 14 emu from D.C.
- Gerber, 14anager, 14P3 Technical Support to 11. Brothers, Director, liillstone Unit 3,14 arch 22, 1995, "RCS Socket Weld Failure".
Subuequently, 14 r.
Pittman stated the present condition was acceptable to start-up from retueling outage ("RFO") 5 and support i
the upcoming operating cycle since an acceptable level of i
confidence existed not to incorporate proposed fixes during RFO 5.
ace, liemo from G.R.
Pittman to 14.H. Brothers, liny 2 0, 1995, "RCS Elbow Tap Socket Weld Failures -- Current Status and Near and Long Term Corrective Strategies".
Furthermore, according to lir.
Pittman, no safety incue existed due to the leak limiting feature of the configuration whereby complete severance could easily be up_by the charg!ng pumps (RCS makeup).
Aiditionally, 14r.
made Pittman stated there was minimal risk.to lost generation.
Basically, lip 3 was willing to live with the problem and put off resolving it for at least one operating cycle even tnough the potentia) for a small break ions of coolant accident ("SBLOCA")
existed due to a questionable dusign.
During the apcoming operating cycle, another failure occurred and was identified during a containment entry to identify RCS leakage foi other sunpected causes.
Ultimately all sixteen locations were modified from the existing design to the current design.
Recent publications validated the shortsightedness of the decision to defer modification by stating tac "AS!4E Code procedure does not ensure the design to be on the conservetive side.
. lioreover, either establishment of a nea design code or confirmation of the authenticity of the current practice, as extended to higher orders of fatigue cycles, appears to be an urgenc need."
See, " Pressure 4
Vessels and Piping Codes and Standards," Vol.1, PVP-Vol. 338 at p.
3.
- But, the bottomline is that this is yet another example of intentional conciu ct on the part of NU to ignore existing safety problens, put ofI needed analysis and repairs, only to have the problem resurface under actual operating conditions.
As previously noted, this in precisely the same conclusion reached in the Executive Summary,141astone Nuclear Power Station, Combined Inspection 245/95-42, 336/95-42 423/95-42, February 6, 1996:
(T]he socket weld f ailures should have been prevented by corrective actions for similar vibration induced failures identified in 1994.
The licensee postponed these socket weld repairs during the 1995 refueling outage, based in the inappropriate safety judgment that any socket weld failures would only result in a cmall RCS. leak.
Id. at p. v.
Since being forced to heighten inspections, by public disclosures 21
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of NU's intentional misconduct and the NRC's prior knowledge of that misconduct, many of the NRC's own inspection reports have validated the contention that NU cannot and will r.at operate within 1
the confines of its licenses, UFSAR's, and Commission regulations.
For example, in a letter to Mr. Feigenbaum, dated July 31, 1996, regarding NRC Inspection Report 50-213/96-201, entitled "Special Inspection of Engineering and Licensing Activities at Haddam Neck,"
the NRC concluded:
The (inspection) team found a number of oignificant deficiencies in the engineering calculations and analyses relied upon to ensure the adequacy of the design of key safety systems at Haddam Neck.
In some cases, design-basis calculations and analyses were not sufficient to conf.irm that the safety system functional requirements would be met.
Some of these errors were longstanding,
. These deficiencies revealed significant weaknesses in the defonso-in-depth princifics that the NRC relies upon to ensuro that nuclear powcr plant operation does not jeopardize the health and safety of t.se public.
The team concluded that weaknesses in your configuration management processes and a lack of technicr1
- rigor, thoroughness, and attention to detail in the design process, either contributed to or directly caused the identified errors.
Id. at p.
1 (emphasic added).
The inspection team also found instances where commitments made to the NRC had not been met by NU.
Id.
Summarizing the report, the NRC noted that NU's failure to act had larger implications for Connecticut Yankee:
The team found several instances involving the failure to identify, evaluate, and correct conditions adverse to quality, and some instances in which planned corrective actions were not promptly initiated.
In some instances, the delays in initiating planned corrective actions were significant because the actions included the evaluation af the potential generic implications aE those iscues for athar plant systoms and equipment.
Id. at p. 2 (emphasis added).
Despite.the fact that NU officials vehemently denied any correlation between its actions at the Millstone units, and indeed pointed to Connecticut Yankee as proof of its commitment to nuclear safety:
The team found process issues at Haddam Neck which are similar to some of those identified at Millstone 1,
as documented in the Event Response Team Report, dated 22 i'
_ _ _ _ _. _ _ ~ _ _. _ _ _ _ _. _. _ _ _ _ _ _ _ _
. o February 22, 1996, that is commonly referred to as ACR 7007.
As discucced
- above, the team found that calculallons did not exisc to support some of the design-1 bases and adminiatrative control programs at Haddam Neck have not maintained an accurate UFShR.
In addition, licensee management oversight did not identify and address the patterns of corrective action program implementation problems, as discussed above.
- Finally, g
ACR 7007 states that a " general lack of understanding and appreciation for the relationship between 10 CFR 50, l
design-bases, licensing-bases, industry codes, and NU's adn.i nis t ra tive programs" existed.
The team observed
[
instances that deruon s t ra ted a
similar lack of understanding by the licensee's staff at fladdam Neck.
l Id. (emphasis added).16 The inspection report itself concluded:
The inspection record for the last 2 years contains instances of problems which indicated weaknesses in the engineering programs and the modification precesses.
1 Instances included cases where inadequate designs were j
installed, engineering analyses were incorrect, or incorrect design-basis assumptions were used.
For i
example, the incorrect design of an auxiliary feedwater flow instrument modification in 1995 resulted in the i
instruments being over-ranged when the main feedwater pumps were running (IR 95-19).
In other examples, nonconservative LPSI flow assumptions were used in LOCA analysis (IR 95-27),
nonconservative assumptions for metal mass were used in the containment analysis (IR 95-1 06), and nonconservative assumptions for the nuclear 4
instrumentation detector response were used in the boron i
dilution analysis (irs 94-09, 94-14).
Although common causes were not identified, these issues indicated a weakness in the processes to verify accurate inputs to i
the licensing-basis analyses.
The inspection report also contains examples where the licensing-or design-basis requirements for some systems were not maintained.
Examples included the failure to meet the design codes for ECCS piping flanges (IR 94-03) and an inadequate seismic interaction analysis for reactor coolant support systems (IR 94-17).
{
inspection Report at p. 53.
The report went on to observe:
16/ Amazingly, although the report noted that the NRC was considering escalated enforcement action, not a single notice of violatina has been issued by the NRC at this time.
23
a The most significant issue noted by the team was the f ailure of the licensee to appropriately consider design-basis scenario loade on the Class IE station batteries sizing calculations.
Specifically, the licensee's calculat. ions did not account for all of the loads associated with a LOCA coincjdent with loss-of-offsite power, and did not demonstrate that the battery voltage would remain above the minimum level required for operation of equipment.
The licensco's evaluation resulted in the station batteries being declared inoperable and major load modifications being subsequently made to reduce loading on the batteries.
Other deficiencies of particular significance identified by the team included failure of the licensee to adequately evaluate the ef fects of two phase SW [ service water) flow at the discharge of the CAR (containment air recirculation) fan coils during accident conditions, errors in engineering analyses and calculations supporting a
proposed TS (technical specification) amendment for lowering CAR air flow, and errors in engineering analyses and calculations for the RWST level instrumentation.
Id. at p.
54.
Other NRC inspection reports support Petitioner's contention that NU'n conduct was deliberate and intentional.
Common sense simply dictates that the number of instances of desigh and licensing-bases discrepancies could not be due to mere negligence alone.
For example, in a letter to Mr. Feigenbaum accompanying NRC Integrated Inspection Report 50-213/96-08, dated September 12, 1996, Mr.
Cooper wrote that NU was expending considerable resources at Connecticut Yankee to addresses such discrepancies "needed to assure the continued operability of components important to safe plant operation "
Id. at p.
1.
For example, Mr. Cooper noted:
The discovery by design engineering that the service water piping supplying cooling water to the CAR fans would not remain f unctional under accident conditions was an example of an issue for which the design basis for the plant had not been thoroughly reviewed or understood.
Other design basis issues discussed in the enclosed inspection report included the reliance of high containment back pressure to assure reliable performance of the residual heat removal (RHR) systems under postulated accident conditions, and the adequacy of the containment sump screens to limit debris from entering the safety systems.
These issues impacted the operability of the emergency core cooling systems from performing their intended safety functions for certain postulated design basis events.
24
s
{
Id.
(emphasis added).
Once again, the NRC noted five apparent violations that were peing considered for
- escalated enforcement" action, but no such action has taken place and no notices of violation have been issued.
Tbe Executive Summary to the inspection report notes a number of instances of inadequate engineering support at tonnecticut Yankee-a theme which appears repeatedly in violations at the Millstone units.
For example, the Summary notes:
[E)ngineering support to operations was inadequate in the failure to provide uncertainty calculations for the wide range nuclear instrumentation trip setpoint, and the failure of the plant design basic to adequately consider two phase flow conditions in the service water supply to the CAR f ans.
The latter deficiency, in combination with other design basis assumptions to assure adequate not positive auction head (NPSH) conditions for safety related pumps, resulted in past plant opera t ion with a potentia 1 for loss of containment huat
- removal, containment integrity and long term core cooling under postulated accident conditions.
Engineering support was inadequate to assure the emergency core cooling system (ECCS) flow path remained operable for all design basis conditions.
Discrepancies included inadequate configuration management of the containment sump design and as-built conditions; a lack of detailed analysis and technical justification for the reliance on post-accident back pressure inside the containment to assure adequate NPSH for the residual heat removal (RHR) pumps, inadequate inspection /varificatiun of sump as-built and material conditions; and, the lack of aggressive action in response to generic communications of indu.:try events, which contributed to an inadequate operability determinacion regarding the cump screen design and mesh size.
Poor engineering support was also noted to assure adequate controls exist for containment isolation valves that might not close in response to an isolation signal, and based on discrepancies in the design basis documentation f or the core delut,e valves.
Id. at pp. 3-4 (emphasis added).
The problems relating to the containment sump pump design serve as a graphic example of the f act that NU's conduct was intentional and deliberote.
As the inspection report noted, since the early 1980's the NRC issued several generic communications to licensees 25
"focon[ing) on the design and po t e n t. i a l vulnerabilities of the containment nump.*
Inspect. ion Report at p.
19.
On 1;ovember 18, 1980, the NRC requented that liU provide information on the design featulen and dimennions of the containment sump.
- Later, on December 3, 1985, the NRC issued Generic Let ter 85-22 indicating it would not require a generic backfit to cor rect the problems.
ll o w e v e r, in 1986, HU reported
- t. h o t i t.
had evaluated containment nump perf ormance in it.n ISAP and considered the matter closed since the emergency nump war. of standard design.
Id. at pp. 19-20.
But:
On July 26,
- 1996, the innpector observed licensee enginecro perf orm a verification of the containment sump nereen menh nizn The nump nereen menh according to in-plant walk down documentation in 1990 indicated that the mech nine had approximately 0.375 inch openings.
The walk down on July 2f, 1996 concluded that the mesh size was generally 0.50 inch, with a 3 inch by 2 foot gap on one end.
Id. at p.
- 19. The NRC inspectors concluded:
The na f e t.y consequences of improper menh in t.h e containment sump is a potential failure t.o mitigate a lons of coolant accident due to debris clogging components within the emergency core cooling systems during long-term recirculation phase.
The ability of operators to transition f rom injection phase to long-term recirculation phane is considered risk significant from the licensee'n individual plant examination (IPE) report.
The vulnerability accounts for approximately 16 percent of the analyzed total core damage frequency.
Id.
at pp.
21-22.
The inspectors found the discrepancy had "nignificent safety consequences based upon plant risk and the ability to mitigate a postulated design basis accident."
Id.
In lat.e August and early September
- 1996, NU's willful and intentional disregard of
- UFSAR, licensing and regulatory requirements over an extendcd period of time led plant operators to place the Connecticut Yankee reactor in a position that nearly resulted in the inabilit.y to maintain adequate decay heat removal to encure the public health and saf et.y.
The bottom line of this event is that NU cont inues to pose a significant risk to the public health and safety whether or not it operates its existing plants.
The event seriously calls into the question the ability of NU to maintain safe conditions even when its plants are in shutdown mode.
In a letter to Mr.
Feigenbaum thct accompanied NRC Augmented Inspection 'leam Review of Undetected Introduction of Nitrogen Gas into the Reactor Vessel during Plant Shutdown, Report No.
50-213/96-08, Region 1 Administrator Hubert J.
Miller stated:
26
i.
.e '
t For approximately four days, control room operators were i
unaware that nitrogen gas was leaking into the reactor I
vessel and causing level to decrease.
By September 1, 1996, reactor vessel level had decreased to approximately l
3 feet below the reactor vessel fic go.
The decrease in reactor vessel level was potentially significant because a f t'rther decrease in level could have challenged the function of the operating heat decay removal system.
While there were not actual public health and safety consequences of this event and adequate decay heat removal was maintained, the situation involving an i
unintended decrease in reactor water level in combination with the unavailability of decay heat removal equipment l
was safety significant.
Several operations proceduros failed to provide adequate details or contained incorrect information. The absence of acceptable procedures was a contributing cause for both the nitrogen gas intrusion going undetected and for the inadvertent diversion of water from the reactor coolant system (RCS).
Several of the events were exacerbated by plant operators failing to follow plant procedures, conducting activities without procedural guidance, or making inappropriate decisions.
A lack of a
questioning attitude resulted in the failure to 4
i promptly identify the nitrogen gas accumulation in the i
reactor vessel.
The failure by more senior operators to convey expectations to less experienced field operators during pre-job briefings resulted in inappropriate equipment manipulation that either directly caused or contributed to these events.
i Id. at p.
1.
The inspection report further found that the " poor material condition of several isolation
- valves, and past 4
^
management failures to address previously raised cc rns about vent header design contributed to the event.
Id.
at p.
2.
Contributing factors also included inadequate procedures, failure to implement procedures, lack of a
questioning
- attitude, inappropriate decision-making, inadequate pre-job briefings, failure to report the event, failure to conduct planned training, avoidable delays due to the RHR pump repair, poor plant material condition, untimely technical response, poor implementation of technical information, weak engineering / operations interface, 2
inappropriate outage scheduling decisions, lack of direct reactor d
vessel indication and slow initiation of event response team.
Inspection Report at pp. 10-16.
In other words, the event was caused by_NU's deliberate and intentional disregard of numerous requirements imposed by Connecticut Yankee's license and UFSAR, technical specifications, and Commission regulations. Still, no notice of violation has been issued by the NRC.
Yet, on November 2,
1996, at Connecticut Yankee an event that 27 i
1
' '. j.'. j L
occurred in the fuel transfer canal and reactor cavity led to the l
unnecessary exposure to workers no airborne radioactive material.
1 The NRC identified iive " apparent " violations, "some with multiple examples of non-compliance, Letter to Mr.
Feigenbaum accompanying NRC Inspection Report 50-213/96-12 at p. 2.
Included 3.1 the findings of the Inspection Report are the follcwing incidents which demonstrate not only intentional and deliberate misconduct, but a callous disregard for the health and safety of its employees:17 On November 2,
1996, two workers in the fuel transfer canal unknowingly collected, handled, and transported radioactive material (debris) with contact radistion levels ranging from 20 R/hr to 60 R/hr.
The debris was not surveyed as it was collected, handled or transported.
Such surveys were necessary and reasonable to ensure conformance with the occupational dose limits.
On November 2, 1996, airborne radioactivity surveys were not _ adequate to detect high concentrations of airborne radioactivity within the fuel transfer canal as workers collected highly radioactive dry dirt like debris therein.
Such surveys were reasonable in that areas traversed and worked in by the workers exhibited loose surface contamination level measuring up to 80 mrad /hr (beta) contamination and up to 30,000 disintegrations per minute /100 square centimeters alpha contamination (dpm/100 cm2).
On November 2, 1996, airborne radioactivity surveys were not adequate to detect high concentrations of airborne radioactivity within the reactor cavity to support reactor stud hole cleaning.
As a result, two woi;kers were permitted to enter the reactor cavity notwithstanding the presence of high levels of airborne radioactivity.
As of November 7, 1996, the licensee had not effectively evaluated the potential exposure of two workers, known to have been exposed to high levels of airborne radioactivity, suf ficient to make the determination that the workers had substantial potential to erreed 17 / This event is remarkable, too, in that it exhibits the same willingness to deliberately jeopardize the health and safety of its workers, that NU exhibited in its refueling practices at Millstone Unit 1.
At Unit 1,
NU sent workers in to begin preparations for moving fuel well in excess of the required hold time.
However, such action did r>t stop NU from publicly arguing the its full core offload practice at Unit I was safer for it.,
Woladrs.
28
m -
4 %'
g I
applicable regulatory limitt relative to intake of alpha emitting isotobes on November 2, 1996.
Inspection Report. at p.
iii.
The report goes on to note numerous other violations in connection with the incident.
In the letter to Mr. Feigenbaum, accompanying the report, the NRC noted:
The NRC inspection identified significant deficiencien in the overnight and control of licensed activitiec, including programnatic breakdc;wn in radiological controls and poor work planning, control, and practices relative to defueling activitleu on November 2,
1996.
As a i
result, personnel were exposed to high concentrations of airborne radioactive material and handled highly radioactive debrin, resulting in a nubstantial potential for an occupational exposure in excess of NRC regulatory limita.
We are particularly concerned about your organization's failure to (1) adhere to fundamental radiological safety requirements (such as offective communication and underotanding of work ocope, knowledge of actual radiological conditions and potential cafety concequence, and conduct of appropriate radiological surveyo or evaluations);
(2) recognize the potential health and snfety concequence of the emergent situation and respond appropriately; and (3) recogniec and ef f ectively communicate to management, a situation which delayed defueling activities and resulted in maintaining the reactor in a heightened shutdown risk condition for an extended period.
Further, we are concerned that your staff failed to recognize that a substantial potential existed 1or pornonnel exposure to altborne radioactivity containing alpha emitterc and conceauently failed to initiate timely and appropriate pornonnel exposuro 0"aluation.
Denpite, the five " apparent" violations, as in the previously noted events, the NRC has yet to issue any notices of violation.
The i ncidents cited above are merely exemplary of the type of conduct in which NU han engaged for the past 20 years at all three Millntone units and at Connecticut Yankee.
The exampleu are not meant to be exhaustive of the evidence in support of the contention that NU han knowingly, willingly and recklessly opereted Millstone Unit 1, Unit 2, Unit 3 at Unterford, CT, and its Connecticut Yankee Nuclear Power Plant at Haddam Neck, CT, in violation of their respective operating licenses, the regulations of the NRC, and their respective UFSAR's for a prolonged period of time which unnecessarily but significantly compromised public health and safety by eroding the required defense in depth philosophy.
In addition to the items specifically cited above, Petitioner requests 29
l s
that the Commionicn take administrative notice of the events reflected in the documents listed la the Section
- entitled, "ldifunce in.Jilipp30rl1Ee11Linn," Lelow and to additional eva.dence that Petitioner plans to present at the requested public hearings.
Content imLNh.]
i The evidence is equally overwhelming that NU has kriowingly,
willingly and intent onally harassed, intimidated and discrimir.ated against its employees who raise r.afety concerns in violation of United St.ates statutes and NRC regulations for a prolonged period of time.
The NRC Office of Investigations has completed one investigation into Petitioner's allegations that hewassubgeted to HILD as a result of raising safety concerns about CU-29.
It currently is investigating HILD allegntione related to LPCI heat exchanger scale.
In addition to the incidents of HI&D to which Petitioner was subject, and which are discussed in more detail below, the NRC itself has acknowledged, albeit belatedly, that NU has a long history of HILD.
In September 1996, the Millstone Independent Review Group ("MIRG") issued its report entitled,
" HANDLING OF EMPLOYEE CONCERNS AND ALLEGATIONS AT MILLSTONE NUCLEAR POWER STATION UNITS 1, 2& 3 FROM 1985 - PRESENT."
The report concluded:
The MIRG determined that in general, an unhealthy work environmeat, which did not tolerate dissenting views, and did not welcome or promote a questioning attitude, has existed at Millstone for at least sevaral years.
This poor environment has resulted in repeated instances of discrimination and ineffective handling c employee l
concerns.
The vast majorit" of employee concerns and alleontions that were submitted at Millstone represented lis ' e safety significance;"
however many involved potentially impo rtan t.
procedural, tagging, or quality 18/ In response to a Iequest under the Freedom of Information Act for all transcripts of Petitioner's interviews with 01, Petitioner was informed that "[t1hese records have been referred to the Department of Justice f or consideration of prosecutive merit. "
Reaponse to Freedom of Information Act (FOIA) Aequest No.96-382, dated October 7, 1996.
It is Petitioner's understanding that OI investigative resulta are not sent to the Department of Justice unless OI has entered a finding of intentional misconduct.
"/ Petitioner disputes that the majority of allegations were of "little sa f ety significar.co. "
Even assuming it is true that most allegations, taken in isolation, were not saf ety significant, the cumulation obviously was of great safety significance as evidenced by the fact that none of the plants currently will be allowed to restart without a vote of the commission.
30
i i
,J d
t casurance (QA)
- problems, and tow were ultimately
%termined to have safety Significance.
The unbenMby w.n k envicor.mont combined with the slanificance of substant. lated allegations contributed to 14111ste ne being I
]
placed on the t!Ec'c votch lit,t in January 199f.
Id. :.t p.
1.20 As to the caures of e.his " unhealthy environment,'
the HIRG concluded:
I i
[T] hat these toot causes underscored a common theme of top management f ailure to provide the dynamic and visibic leadership needed to brim; about required, basic attitude changes.
!Jone of the findinqu of this teau are tr ew.
Every problem identified during this review had tieet.
pr<wiously identified to 140 management, often by its own self-acueriuments, yet the same problems continue.
This single failure la viewed as be bnj at the core of M131chne's continuing employee concerns.
Id. at p.
2.
This root cause, standing alane, de".onetrates that 1;U, an a corporate entity, deliberately and intentionally created the hostiJe work environment and then allowed it to persist.
4 Petitioner has been the subject of long-standing and pervasive hit.D.
In 1991, Petitioner was the immediate supervisor of 14r.
Galatis when he firnt started pursuing internally NU's full L.oro offloading practico at 14illstone Unit 1.
Petit.ione:
Wo essentially the first person in the chain of management to agree 1
- t. hat Mr. Galatis wac correct in his assertion that M111 stone Unit I was in violation of its license, At the same time that he was supporting Mr. Gulatis, Petitioner also was pursuing, internally with NU, his concerns about ISAP and CU-29.
In December 1993, Petitioner was demoted from his position of Engineering Supervisor after supporting Mr.
Galatis and completing a formal evaluation of the CU-29 ist,ue in REF 92-84, which NU management repeatedly rejected but Petitioner ultimately reversed.
/ It chould be noted that the harassment of Mrs Galatis and 20 Petitioner specifically were excluded from the MIRG study, as was that of several other individuals, including 104 employees who were subject to layoff in January 1996.
The NRC, in conjunction with the Department of Labor, conducted an inquiry into the January 1996 layoffs, which is yet to be made public.
However, the Department of Labor has concluded that all of the 104 severance agreements 4
contained a clause which was void as against public policy because it required the employees to waive the rights to file a complaint under the Energy Reorganization Act or fc-feit their severance benefits.
31
-,a s e s
After his demotion, "otitioner expnrionced a decline jn his performance t atings as he continued to pursue t.dety-related issues both internally and with the NRC.
]n tne summer of
- 1995, Petitioner was not tolectud for a technical position with Public dervice Co. of New Hampshire ("PSNH") which s af filj uted witn NU's Seabrook Nuclear Power Station.
In November 1995, Petitioner was given a psych, logical evaluation au part of an application for another nupervisory position at NU.
In December 1995, Petitioner was given his results.
Petitioner was adviced that his evaluation was unacceptable in eight of 11 categories.
Typical of HI&D, Petitioner was advised of incredibly poor interpersonal skills, inferring he was nothing but a liability a
to NU.
NU withhold the written report of the evaluation which Petitioner was promised.
From at least 1993 on, Petitioner has been continually subjected to a hontile environment which was deliberately and intentionally created by NU management in order to discourage employees from identifying and pursuing safety issues.
Finally, in August 1996, Petitioner transferred from his engineering position to a senior analyst ponition in the Employees Concerns Prograu ("ECP") in hope of extracting himself from the hostile environment.
Petitioner contirioes to be subjected to a hostile environment in his present position by limited partacipation opportunitier for his qualifications and experience.
4 In f urther support of this contention, Petitioner requests that the NRC take adminis t.rativo notice of all documents, transcripts,
' reports and findings f rom OI investigations and Department of Labor proceedinJ6 involving Eliot Abolofia, Dnnald Del Core, Sr., Timothy O'Sull m n, Paul Blanc!., C+arga Galatin, George Betancourt and Harry scully.
Contention _Nh_3 Dased on all of the above, as well as the evidence incorporated by reference, D?fra, it a clear that NU cannot provide reasonable assurances hat it cao fease and desist. from engaging in similar act ivities in the future.
To borrow a commonly used cliche, "the promises (of NU) are not worth the paper they are written on. "
And, the protection of the public health and safety demands more than paper promises.
The protection of the public health and safety demands that a clear set of operating conditions be placed on Millntone Units 1,
2 and 3 and Connecticut Yankee with clear penalties to be extracted in the event that thc*ce conditions are not met.
Petitioner is by no means the only one to conclude that NU is not capable of properly operating its nuclear power plants.
In an audit conducted for the Connecticut Department of Public Utility Control
("DPUC") by the Barrington-Wellesley Group, Inc.,
the 32
- _ _ _ ~
,,0..<,
t 4
auditors ccacluded:
Heither NU's senior executives nor itu Board of Trustees (Board) have exhibited the leadership and vision necennary to address the fundamental needs of a well-perivrming nuclear program.
NU and its Board had sufficient information to realize that decisive action was necessary to address the deteriorating performance in NU's nuclear operation, especially at MP1 and MP2.
1 NU does not. have a coherent, lonq term strategy for its nuclear operations.
See, "A Focused Management Audit of the Connecticut Light and Power Company's Nuclear Operations," August 30, 1996, Executive Summary at p.
2.
s Similarly, another DPUC audit conducted by R.C. Brown & Associates, Inc., concludedt RCB&A's review of the historical record, as well as the results of interviews with NU personnel, indicated that the issues cited in designating the Millstone site to J to Watch List have been matters of discuccion between NU end the NRC for a number of years, in some cases dating back to the late 1980s.
To address such NRC concerns, NU developed a number of improvement initiatives over the years.
However, such initiatives suf f ered f rom a lack of management direction and commitment, with the result that such initiatives were generally inef fective in addressing and resolving the NRC's concerns.
Between 1992 and
- 1994, NU undertook a
Performance Enhancement Program
(" PEP"), consisting of 42 " Action Plans" designed to address weaknesses in management practices; programs and processes; and performance assessment.
- However, the PEP suffered from lack of direction and support form (sic) senior management, and was terminated without effective validation that the program had achieved its intended results.
In late 1993, following a significant safety-related incident at Millstone Unit 2, NU instituted a significant reorganization at the Millstone site, including the appointment of a new Senior Vice-President, Millstone Station.
The new Senior V.P.
established his own corrective action initiative, designated the Improving Station Performance ("ISP") program.
The ISP featured initiatives in the areas of effective employee communications; leadership development; effective corrective action; procedure development and use; work planning and control; and operations.
As was the case 33 1
i
( ',,, 0.. o
'o *
[I l
with the PEP, the ISP ouffered from a lack of support f rom cenior management, f ailed to meet its most important i
objectives and, ac of early 1996, had been largely abandoned.
)
l See, "Focuned Audit of the Connecticut Li g h t. and Power Companyt Nuclear Operations,"
R.C.
Brown & Annociat en,
Inc., December 31, 1996, at pp v-13 to v-14.
)
A t. a meeting of the Connecticut Nuclear Energy Advisory Committee
("NEAC") on February 20, 1997, an NU official stated that Millstone would try to rentort three unite within one year-a feat never accompliched by anyone in the world.
Of cource, NU is quite familiar with achievemento not previously accomplished.
It 10 indeed an elite group that, in the cource of one year, is denied permincion to rentart four unito.
But, one can only question the abcolute folly of even considering the prospect that NU would be allowed to restart all three unito in a one year period of time.
Yet, that in what the NRC is contemplating even at thic moment, while it allown the previous minconduct of NU to go unpuniched.
Nor in the public'n confidence in t he abilit.y of NU to ref orm it.o wayo in the future rectored by the fact that in December 1996 cix of seven NU employeen who applied for a license to operate Millntone Unit 1 failed the examination.
The Initial Operator Licenning Examination Report, dated February 7, 1997 concluded:
The applicanto were poorly prepared f or the examination, and an a
- result, 6
of 7
app 11 canto failed the examination.
One SRO instant applicant passed by the examination by a very cmall margin.
Significant generic weaknennen were noted in both the writ. ten and operating test, and f our of the neven applicants had written scorea of 72 or below.
Significant weaknennec were identified during the operating teat ej ated to operating reactivity controlo, diagnonic of inst.rument. f ailures, and diagnonic of ECCS injection status.
Equally, NU la poorly prepared for the " tent" ahead of it.
Buc thin 10 not a " test."
Plano on paper, reorganizations, and new management initiativen have been tried in the past.
All have had the same result.
They merely have provided NU with a guine to go about business in ita usual manner elevating cost over cafety.
There ic absolutely no reason to believe that thia time will be any different.
The only thing that can make this time dif forent is to clearly spell out the rules of the game, and impose stiff and swift penaltien in the event the rules are violated.
This is the only manner in which the NRC can give the public nome reasonable assurance that NU will not repeat its pact behavior.
34
o Py EYlde ncp_in_S@pprt_of_Ectition In addition to the evidence specifically cited above, a large body of factual evidence, on which this petition is based, is currently in t-xistence and in the possession of the In c. Incorporated, by reference, into the factual basis for this petition are the following documents:
1.
Petition uubmitted pursuant to 10 C.F.R.
% 2.206 on behalf of George J.
Galatis and We the People, Inc., of the United States, dated August 21, 1995; 2.
Supplemental petition submitted pursuant to 10 C.F.R.
% 2.206 on behalf of George J.
Galatic and We the
- People, Inc.,
of the United States, dated August 28, 1995; 3.
Report of U.S. Nuclear Regulatory Commission, Of fice of Inspector General, "NRC FAILURE TO ADEQUATELY REGULATE
- MILLSTONE UNIT 1," Case No.95-771, dated December 21, 1995; 4.
Report of U.S. Nuclear Regulatory Commission, Of fice of Inspector General, "NRC STAFF ACTIONS TO ADDRESS NORTHEAST UTILITIES SYSTEM (NU) 1991 SELF-ASSESSMENTS,"
Case No.96-02S, dated May 31, 1996; 5.
Report of U.S. Nuclear Regulatory Commission, Of fice of Inspector General, "NRC HANDLING OF ISSUES RELATED TO REFUELING OPERATIONS AT MILLSTONE UNIT 1,"
Case No.96-05S, dated July 23, 1996; 6.
Report of U.S. Nuclear Regulatory Commission, Of fice of Inspector General, "NRC STAFF 4CTIONS TO ADDRESS CU-29 ISOLATION VALVE ISSUE, " Cace No.96-06S, dated September 3,
1996; 7.
Report (s) of U.S.
Nuclear Regulatory Commission, Office of Investigations, regarding allegations of intentional wrongdoing by George J.
- Galatic, date(s) uncertain; 8.
Report (s) of U.S.
Nuclear Regulatory Commission, Office of Investigations, regarding allegations of intentional wrongdoing by Albert A.
- Cizek, date(s) uncertain; 9.
Report (s) of U.S.
Nuclear Regulatory Commission, Office of Investigations, regarding allegations of intentional wrongdoing by George Betancourt, date(s) uncertain; 35 i
- /eU 10.
Millc+.one Unit No.
3, Safety Systems Functional Inspection of the Condencate/Feedwn*er\\Feedwater Coolant Injection System, December 10. 19P8; 11.
NE&O Performance Task Group Report, NU, September 1991; 12.
I'ina1 Report of the llo_ctsinre Complittitcf/_T31Ah_fRLcf2 at Millutone Point Station, date uncertain; 13.
Evaluation of Millctone Nuclear Power
- Station, Inctitute of Nuclear Power Operations, December 1, 1992; 14.
Report of Self-Ascecoment Tank Force, NU, May 1994; 15.
Millotone Independent Review Group, 11a ndling of Employee Concerno and Allegations at Millstone Nuclear Power Station, Units 1,
2 & 3 from 1985
- Present, September 1996.
Colmlusion For the reasons cited f ully above, Petitioner espectf ully requests that the NRC inctitute a proceeding purcuant to 10 C.F.R.
k 2.202 to modif the operating licensen of Millstone Units 1, 2 and 3 and Connecticut Yankee ao described above.
Recpectfully cubmitted7
'/
Ern/ 6l f.*,fi diey est n/
cc: A Cirek
!!. lle l l, IG D.
Kenyon, Prec., NU 36
)