ML20062F922
| ML20062F922 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 11/30/1978 |
| From: | Like I SUFFOLK COUNTY, NY |
| To: | |
| References | |
| NUDOCS 7812210411 | |
| Download: ML20062F922 (145) | |
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UNITED STATES OF AMERICA J
NUCLEAR REGULATORY COMMISSION SEFORE THE ATOMIC SAFETY AND LICENSING SOARD U1 In the Matter of
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LONG ISLAND LIGHTING COMPANY
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Docket No. 50-322
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[{f g 70 (Shoreham Nuclear Power Station,
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Unit 1)
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COUNTY OF SUFFOLK'S PARTICULARIZED CONTENTIONS This submission reviews the status of all contengions raised by the County of Suffolk in these proceedings, providing the technical basis and appropriate reference docunents for each such contention.
It should be noted that the fors.al dis-covery period in this case has not yet begun to run but wi1E be triggered upon issuance of NRC Staff's Safety Evaluation
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Report.
See " Order Relative to Requests for Clarification and N
Reconsideration of the Scard Order of January 27, 1978", p. 5, (3/18/78).
Thus subsequent discovery efforts will likely pro-duce further basis for the County's contentions.
However, it is the County's position that this docunent, together with its previously-filed. pleadings, provide nore than angle basis for full adnission of the County's contentions under 10 CFR, 22.712(a).*
The County, in previous subnissiens, objected to the Hearing Board exclusions of Cententions 2a, fc, 19a, 22-26.
It renews these objections here.
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The County has elected to submit this pleading at this time for several discrete reasons.
As this Board has been made aware, earlier this year the Suffolk County Legis-
.f lature terminated its contract with its consultants for this i
case, MHB Technical Associates.
However, before the Legislature moved to end its relationship with MHB, the Company had embarked upon extensive informal discovery with NRC Staff.
The purpose of this effort was to cut through the formal discovery process and to produce a set of well-defined contentions that wculd facilitate the litigation of this case and produce a concise -
hearing record.
The Legislature, recognizing the benefits of this approach, and the need to bridge the transition of the case to new consultants, earmarked an additional sum to allow MH3 to complete its work and prepare this document.
Staff submitted extensive interrogatories to the County at a point in time when it was still unclear if MH3 would be authorized to continue its work and to utilize the information
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gathered during the informal discovery period.
Because MHB was able to produce the desired document, the County believes that the need for particularized response to Staff's inter-rogatories has now been eliminated.
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,n CONTENTION 2a EXPERIMENTAL SAFETY RESEARCH PROGRAM 1.
STATEMENT OF CONTENTION l
2a.-
Intervenor contends that the Applicant and the Regula-tory Staff have not conducted an adequate design verification experimental safety research program to demonstrate that the Shoreham Plant emergency systems will function in accord with the General Design Criteria as required by GDC 35.
Intervenor also contends that failure to perform adequate design verifi-cation is contrary to the requirements of 10 CFR Part 50, Appendix B, Criteria III and XI.
Such deficiencies specifically pertain to the functionability of the emergency core cooling i
system under all transient and accident conditions.
I 2.
CONTENTION CHRONOLOGY This contention, along with all other currently active County contentions, was originally described in County's Amended Petition to Intervene, dated September 16, 1977.
In the October 11, 1977 Prehearing Conference, it was ruled as one in which the Board would hear additional arguments before making a ruling on admissibility.
Accordingly, contention wording was revised to that found listed above by County's l
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Amended Contentions and Legal Arguments Submitted in Compliance nm.
With Hearing Board's Request of October 11, 1977, dated Nove=ber 10, 1977.
Subsequently, the Board's January 27, 1978 order disallowed this contention ansi the Board's March' 8,1978 order further ruled that this contention would be unacceptable for litigation.
3.
BASIS FOR CONTENTION Basis for this contention was well describec
.a County's November 10, 1977 above referenced document.
Substantial tech-nical uncertainty still appears to exist in this area.
This 2-1
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B is evidenced by the fact that the General Electric Company is in the process of building a large core spray test facility i
at their Lynn, Massachusetts plant.
However, since this con-tention has been ruled inadmissible, further details of its basis will be omitted.
4.
REFERENCES See County's November 10, 1977 filing, pages 2 through 4 for further specific detailed information.
5.
SUMMARY
AND CONCLUSIONS Since this contention has been twice ruled inadmissible, no further technical discussion of it will be included in this report.
O t
2-2
CONTENTION 3a DEFINITIVE SAFETY FINDING - UNRESOLVED SAFETY ISSUES 1.
STATEMENT OF CONTENTION 3a.
Intervenor contends that the Applicant and Regulator Staff have not adequately assessed the impact of numerous,y unresolved safety items, both singularly and cumulatively, in evaluating and reviewing the Shoreham nuclear plant in conjunction with the operating license application.
As a result of this inadequate assessment, the Shoreham systems, structures, and components have not been backfitted to current regulatory requirements in adequate compliance with 10 CFR Part 50.109 with regard to:
1.
The list of generic light water safety items s
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developed by the Advisory Con:mittee on Reactor Safeguards as documented in the February 24, 1977 report entitled, " Status of Generic Items Relating to Light Water Reactors :
Report No.
5."
11.
The list of generic technical activities under consideratio. by the NRC Staff and sumarized in the report entitled, "NRC Technical Safety Activities Report."
iii.
The list of unresolved BWR safety issues discussed by General Electric in the proprietary General Electric Report, " Reed Task Force Report."
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2.
CONTENTION CHRONOLCGY
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This contention, along with all other currently active County contentions, was originally described in County's Amended Petition to Intervene, dated September 16, 1977.
In the October 11, 1977 Prehearing Conference, it was ruled as one in which the Board would hear additional argn=ents before making a ruling on ad=issibility.
In accordance with the Board's request, additional argument was submitted by County in Ccunty's November 10, 1977 A= ended Cententions and Legal Arguments, which formulated the contention wording given above.
Following rec.eipt 3-1
O of this information, the Board ruled on January 27, 1978 that Contention 3a would be disallowed.
County's February 17, 1978 Objections to Board's Order were filed and the Board's subsequent March -8,1978 order admitted Contention 3a as acceptable for discovery to the extent that County could point to specific items and establish a nexus between each item in the Shoreham facility, t
3.
BASIS FOR CONTENTION The basis for this contention is as stated in the September 1977 Amended Petition to Intervene and as described in greate detail in the following.
Detailed assessment'of generic
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issues at this time, is hindered since the Staff has still not issued the Shoreham Safety Evaluation Report (SER).
The term " unresolved technical issues" identifies generic deficiencies which may contribute, perhaps gravely, to ampli-fying the risk to public health and safety from power reactor operation.
In addition, the public through their participation as electric utility rate payers and stockholders is also faced, with assuming the billions of dollars economic burden which may lie in wait due to backfitting of operating reactors and changes during construction'tc planned nuclear facilities following resolution of the " generic issrTs."
Quantification of the potential technical and economic risk, and the associated uncertainties, has remained an elusive goal.
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Evidence of possible inadequacies of U.S. light water reactors is accumulating in the form of " unresolved issues" t
which may affect many operating or planned reactors.
For i
example, 133 issues have recently been designated by the NRC as requiring Staff attention. (1)
These issues, each a t
I potential deficiency affecting many commercial nuclear plants, i
have previsculy been identified in a variety of documents i
including the NRC's Technical Safetv Activities Report (TSAR). (2) the Advisory Committee on Reactor Safeguards (ACRS) lise of l
generic issuesj3) an internal assessment of the technical and
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business risk facing General Electric in fulfilling its orders
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forBWR's,entitledtheReedTaskForceRecorth) and finally, the list of issues (5)(6) raised by concerned NRC Staff members in response to the new NRC procedures for dealing with Staff dissent.
l 1.
ACRS List of Generi'c Items
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l The NRC's Advisory Co=mittee on Reactor Safeguards j
has been publishing since December 18, 1972, a series of lists b-of generic items (problems) which are of concern to light water reactors.
The November 13,,1977 list includes a total of 76 itemsofwhich28areconsideredunresolved{7)
Items listed l
I include important and safety related structures and equipment such as containment, pressure vessel, ECCS components, piping, f
i and electrical supplies.
In Table 1.5.1-1 the Applicant pro-vides a brief summary on the degree of compliance of Shorehan
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to the ACES conce:r.s through 1973.
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The current list of unresolved ACRS issues, as well as a designation of those relevant to a 3WR, is shown in this report in Table 3-1.
Ten of the problems have been on the list since its inception in 1972, yet remain unresolved.
Even when items are declared resolved, there is no certainty that a solution has been implemented.
For example, it is extremely surprising to learn that the ACRS definition of
" resolved" means(0) "in some cases an item has been resolved in an administrative sense, recognizing that technical evalua-tion and satisfactory implementation are yet to be completed."
Anticipated Transients Without Scram (ATWS) represents an example of this category.
In other instances, the resolution of ACRS issues (0) "has been accomplished in a narrow or specific sense, recognizing that further steps are desirable, as practi-cal, or that different aspects of the problem require further investigation."
Examples in this category include the possi-bility ef improved methods of locating leaks in the primary system, and of i= proved methods or augmented scope to inservice inspection of reactor pressure vessels.
Forty-eight " resolved" items exist; it is not known what percentage of them resulted
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in real fixes in hardware or structures and, conversely, how many were " administrative" only.
ACRS review of the Shoreham operating license request is expected during the first half of 1979.
The County, at that time, may wish to present testi-many to the ACRS as related to the County's contentions.
As
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a mini =um, the County should monitor the progress of the ACRS deliberations on Shoreham.
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TABLE 3-1 O}
ACRS CEF:ERic 1SSIES-8L"50f.trrT0*! PENDINC PRIO!tITY FOR RF. LEVANT TO PESOLUTIC3 ACRS CENERic ITt g
Billt ACTG g
CRout' II:(Resolution Pending Since December IS.1972)
- 1. Turbine Missiles 1
X A
A-37,A-32
- 2. Containment Spesys X
3 C-10
- 3. Pressure vessel Failure 3y Thermal Shock X A
A-11
- 4. Instruments to Detect (Severe) Fuel Tailure X
X C
SA. Excessive Vibration,
'I X
B
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- 55. Loose Parts Monitoring I
Z B
3-60 l
- 6. Ison-Random Multiple Failures X
X A
C-13 i
6A. 3eactor Scram Systa=s X
X A
A-9
- 63. Alternatina Current sources Onsite er A-35.3-56, Offsite Z
X A
3-57 i
6C. Direc. Current Systems I
X A
A-30
- 7. Schavior of 8.eactor Fuels Under Abnormal Conditions X
X A
3-22 k ',
l
N 3-68 4
1
- 9. Seismic Sersa Z
I C
D-1
- 10. ECCS Capability for Tucure Plants I
X A
D-2
. CROUP II A (Resolution Pending Since February 13, 1974) l
- 1. Ice condenser contM n-ents I
B B-54 4
- 2. PWR Pump Cverspeed During a IACA Z
B 3-68
- 3. Steam Ge=crator Tube Leakage I
A A-3,A-4, A-5
- 4. ACRS/NRC Periodic 10-year Review I
I C
Policy CROUP TT 3: (Resolution Pending Since March 12, 1975)
I
- 1. Cos'puter Reactor Protection System X
3 A-19
- 2. Qualification of New Tuci Geometry X
I C
3-22
- 3. BUR Mark III contrinnents I
5 A-39,3-10
- 4. Stress corrosion Cracking in SWR Piping I
S Policy CROUP TT Ca (2es41ution Pending Since April 16, 1976)
- 1. Locking out o'f EC::S Power '2perated valves X
X 3
3-8
- 2. resign Features to control Sabacsse X
X
'A A-23
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3A. Decontactination Z
I E*
A-15
- 33. Decocruissioning I
X 3
3-64
- 4. Vessel Support Structures X
B A-2
- 5. Water Macraer I
X A
A-1
- 6. Maintenance and Inspection X
X 3
3-24
- 7. SWR M. ark I Contatncents X
A A-6.A-7, A-39 i
C: tot:P t! 3: (Resolutten Pending Stnec February If.,
19*7)
- 1. Interfaces X
X A
- Policy, A-17
. Capabilit.y nf iter etic Sest's X
X C
C-7.
C:*ntif' t t R :
(R. enlut ton Pendtnt, Sincu flove:nlacr 15, 1977)
- 1. Soil-struettare Interact!an X
X C
A-40,A-41 3-5
ii.
GE " Reed Recort" j
The Reed Report provides a significant assessment of the expected safety and availability for GE-designed nuclear power stations.
The in-depth critical examination of all aspects of General Electric Boiling Water Reactors (GE is providing the Nuclear Steam Supply System for the Shoreham Station) was re-vealed by Reginald Jones, Chief Executive Officer of the General Electric Company, in a December 1975 address to the New York Security Analysts.
Jones stated that in late-1974, Dr. Charles Reed, a top technologist at GE, amassed a task force of the most knowledgeable people that could be put together in the nuclea:. y business.
This task force, including as r.any as 70 to 80 people, worked for a whole year.
It included the finest scient3.sts and engineers in GE.
The result was a final report that was over-whelming---a five-foot shelf full (9)
The Reed Task Force Report (often called the " Reed Report") identified over 100 items required to improve the performance of GE BWR's, of which 27 items were also determined by GE to be safety-related(10)
Of course, almost all the items identified by Reed may impact on plant operating benefits with regard to plant down time, addi-tional maintenance, and equipment changes and modifications.
Dr. Reed described the purpose and objectives of the Reed Report by quoting the opening paragraph of that report:
" Objective of Study'. The Unclear Reactor Study was a highly technical s tudy with the obj ectives of determining the basic recuirements for i= ale-menting the Nuclear Energy Division's (NED) quality strategy through continuing improvement in the availability and capability of Soiling Water Reactor Nuclear Plants (3NR's).
This strategy is predicated-3-6
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on the view that leadership of the BWR in 'these characteristics represents the greatest opportunity for reducing the Utility customer's power generation cost, with resulting lower power cost for industry and for the ultimate consuming public.
The s tudy included review of the broad range of opportunities for development of BUR leadership in all aspects of availability and capability across the entire range of design, development, manufacturing, cons truction and op.eration." (11)
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Dr. Reed elaborated on the purpose and objectives as follows:
"The task force made nu=erous reco==endations intended to i= prove the availability level of g.
h-the BWR.
These recommendations dealt with over-l i
s all reactor design considerations, as well as with specific plant components and services.
We also made recommendations concerning development and test facilities, and concerning questions of.
management and' organization.
The report is, of course, a docu=ent of considerable' sensitivity from a competitive standpoint.
It candidly dis-cusses opportunities for improvement in our product line and our organization ~ and reevmmends steps to strengthen our competitive position." (12)
In February 1976, in response to questions from Congress, the
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NRC Staff reviewed the Reed Report with the purpose of deter-
) CL mining. whether all items of safety significance identified in the report had been reported to the NRC as required.by Section 206 of the Enerzy Reorganication Act of 1974 GE did no t
' bublicly release the report since they believed that-public identification of such a critical self-examination could have l
an adverse effect on the competitive position of GE.
Therefore, l
i two senior members of the NRC Staff who were fc=iliar with the i
design and related safety issues of GE reactors, spent nearly l
two days reviewing the study at the Washingten, DC offices of 1
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of GE in the presence of the GE manager of. licensing.
The NRC concluded in part that:
"In our review of the GE nuclear reactor study, it I
was apparent that the study was mainly directed
. at the marketing rather than safety per se.
The report does contain items which had implications-on the safe construction and operation of BWR's; however, the examples were used to illustrate the point that identified problems (some of which had safety significance) do have an effect on che availability of BUR plants and hence the cose and marketing potential of that plant.
In those instances where probicms having safety significance were cited there was no analysis in the GE report
'of the significance from a safety standpoint of the particular phenomena." (13)
The NRC did not require that the "eed Report be submitted to
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the NRC, or that a list of the safety-related issues contained therein be submitted (10)
Two years later, in February, 1978, in response to a letter from Congressman Dingell concerning the present status of the 27 safety issues, NRC Chairman Hendrie responded that;
'.' Notes taken by the ' staff members during the reb.ew of the study listed the safety issues identified (V
T by the Staff and included the 27 issues identified by the GE licensing group.
However, the notes taken
'during the S taff review were never formally cocu-mentec ano cannot be locatec.
Therefore, we will request GE to release the Reed Report or at least the list of 27 safety issues they identified.
If this proves unsuccessful, members of the Staff familiar with the report and'our ongoing program to resolve outstanding generic issues, will meet with GE to verify and document that all safety issues. identified are being adequately addres'ses."
(emphasis added) (13) 3-8
Since the Shoreham plant is a GE BWR of the type assessed in the Reed Report, the 27 safety items noted in the Reed Re-port may all or in part apply to provisions of the Shoreham design.
Thus, information concerning the NRC review of the Reed Report, and specific information concerning the identification and status of the 27 safety items should be made available to the County and to the Atomic Safety and Licensing Board (Board)(16) so as to permit a complete and thorough safety assessment of the plant.
In contrast, on July 10, 1978, the NRC determined f
that "public disclosure of the aggregate list of the 27 issues a
could cause substantial harm to the competitive position of GE."(17)
The NRC apparently gave greater weight to GE's marketability of BWR's as contrasted to the full public disclosure of known, but unresolved items.
Further= ore, in response to a subpoena for the complete Reed Report by the Board in the Black Fox Unit construction permit proceeding, General Elect-ic in part responded that:
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".....Section 206 of the Energy Reorganization Act C
of 1974 and 10 CFR Part 21, the'NRC Regulations implementing that statute, obligate directors or responsible officers of firms engaged in supplying nuclear equipment to report any defects or items of noncompliance which relate to a substantial safety hazard.
This "section 206 review" did not attempt to define every matter discussed in the Reed Report which might arguably relate to safety.
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The standards contained in 10 CFR Part 21 and Section 206 contemplate a higher threshold to i
trigger a reporting obligation than a mere re-lationship to safe:y.
Thus, the 27 issues which
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were preliminarily identified by GE pursuant to this review were reviewed agains t the more strin-gent standards arising from Section 206, and did not necessarily include all matters discussed in the Reed Recort which might arruably relate to safetv.
(l'S)
(emphasis accec) i 3-9 l
In summary, the preceding GE response indicates that the
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entire Reed Report must be reviewed since more than the 27 issues "might arguably relate to safety."
9 iii.
NRC Generic Safety Program Unfortunately, the ACRS list and the Reed Report findings are by no means a complete accounting of " unresolved issues."
The NRC in 1974 tabulated the unresolved safety issues into an internal Technical'S'afety Activitie's Report, but did not commu-nicate to the public the requirements for resolution.
The TSAR in December,1974 listed 223 items of concern to the NRC of f
.n which 173 of the items were categorized as having "an importaht impact on the licensing review process."(19)
In response to this long list of unresolved issues, the slowness of the NRC in addressing these issues, and as a result of Congressional action on the Nuclear Regulatory Commission budget for Fiscal Year 1978, the Energy Reorganization Act of 1974 was amended (PL 95-209) to include a new Section 210 which reads as follows:
" UNRESOLVED SAFETY ISSUES PLAN" "SE C. 210.
The Commission shall develop a plan
- providing for specification and analysis of unresolved safety issues relating. to nuclear reactors and shall take such acticn as may be
- necessary to implement corrective measures with respect to such issues.
Such plan shall be sub-
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mitted to the Congress'on or before Janua y 1, 1978 and progress reports shall be included in the annual report of the Commission thereafter."(20) 3-10
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.r The joint Explanatory Statement of the House-Senate Conference Committee for Bill S.ll31 provided additional information
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regardine its deliberations on this portion of the bill.
The conferees expressed the intent that the NRC plan should identify and describe those safety issues, relating to nuclear power reactors, which are unresolved on the date of enactment.
The plan should set forth:
(1) NRC actions taken directly or indirectly to develop and implement corrective measures ; (2) future actions planned concerning such measures; and (3). time-tables and cost estimates of such actions.
The NRC was also
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requested to indicate the priority it has assigned each issue, and the basis on which priorities have been assigned.(21)
Implenentation of the NRC program began in April,1977.
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Initially, each of the four NRC divisions reporting to the Office of Nuclear Reactor Regulation described and proposed i
those generic items it considered to warrant the highest.pri-ority.
Proposals were received for 355 tasks (22)of which, i
following consolidation and elimination,133 tasks were eventu-(;
ally selected for review.
A set of uniform criteria (shown in Table 3-2) for generic technical activities indicative of their priority for resolution was developed.
The 41 tasks shown in Table'3-3 were classified as Category A.
As indicated I
in Table 3-3, the remaining issues were ' divided into 72 Category j
B tasks,17 Category C tasks, and 3 Category D tasks.
The definition of Category A issues is strikingly similar to the definition in the NRC regulations of issues which may require i
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TABLE 3-2 i
PRIORITY CATEGORY DEFINITIONS, Category A:
Those generic technical activitics judged by the staff to warrant priority attention in terms of manpower and/or funds to attain early resolution.
These matters include those the resolution of which could (1) provide a significant increase in assurance of the health and safety of the public, or (2) have a significant impact upon the reac, tor licensing process.
Category 8:
' Those generic technical activities judged 'y the staff to be important in o
assuring the continued health and safety of the public but for which early resolution is not required or for which the staff perceives a lesser safety, safeguards or environmental significance than Category A matters.
Category C:
Those generic' technical activif.ies judged by the staff to have little direct or icmediate safety, safeguards or environmental significance, but which could lead to improved staff understanding of particular technical issues or refinements in the licensing process.
T I-Category D:
Those proposed generic technical activities judged by the staff not to warrant the expenditure of manpower or funds because little or no importance 'T to the safety, environmental or safeguards aspects of nuclear reactors or V
to improving the licensing process can be attributed to the activity.
- Appendix B of NUREG-0410.
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l "backficting" of structures, systems, or components of a nuclear facility.
NRC regulations state that backfitting is appropriate if "such action will provide substan'tial protec-tion.....for the public health and safety."{ 3)
(emphasis added)
In a like manner, the Category A issues are those matters which in part "could provide a 'significant increase in assurance of the health and safety of the public."(24)
Based on the similarity of the preceding definitions, it is clear that, as a minimum, all relevant Category A issues could potentially i= pact the ' safety assessment of the Shoreham Station.
Likewise, a review of the Category B, C, and D items may reveal one or more that are particularly troublesome for Shoreham and should, therefore,. be backfitted.
Detailed Task Action Plans have been initiated by the NRC for all Category A generic issues.
Task Action Plans include a description of the problem, the Staff's approach to 1ts resolution, the NRC technical organizations involved in the review and estimates of the manpcwer required from each, a description of the inter-V actions with other NRC offices, the Advisory Co=mittee on Reactor Safeguards and outside 'organi:Ations, an estimate of any funding required for contractor-supplied technical assis-tance, a schedule for completing the task, and a description of any potential problems that could impact the plan.
Task Action Plans have not yet been publically disseminated for the 92 issues in Categories 3, C, and D.(25)
The NRC program to assess the risk associated with the generic issues is not complete as the Staff candidly acknculedged in their following testimony in the Black Fox proceeding:
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7, TABLE 3-3 CATECORY A TECHNICAL ACTIVITIES RELEVANT TO TASX NO.
TITLE:
PWR 3WR A-1 Water Ha=mer X
X-A-2 Asymet'ric Blowdown Loads on the Reactor Vessel X
X A-3 Westinghouse Steam Generator Tube Integrity X
A-4 Combustion Engineering Steam Generator Tube Integrity X
A-5 Babcock & Wilcox Steam Generator Tube Integrity X
A-6 Mark I Short Term Program X
A-7 Mark I L ong Term Program X
A-8 Mark II Program X
X A-9 ATWS X
X A-10 BWR Nozzle Cracking X
A-ll Reactor Vessel Materials Toughness X X
A-12 Fracture Toughness of Steam Gener-ator and Reactor Coolant Pu=p q
Supports X
X i
)
A-13 Snubbers X
X s
A-14 Flaw Dstection X
X A-15 Decontamination I
I A-16 Steam Effects on BWR Core Spray Distribution X
A-17 Systems Interaction in Nuclear Power Plants X
X A-18*
Pipe Rupture Design Criteria X
X A-19 Digital Co=puter Protection Systems Plants with digital ecmputers.
A-20 I= pacts of Coal Fuel Cycle Environmental A-21 Main Steam Line Break Inside Contain=ent X
A-22 PWR Main Steam Line Break - Core and Pri=ary Coolant Boundary Response (MSL3 Outside Contain=ent)
X A-23 Contain=ent Leak Testing I
X Qualificaticn of Class IE Safety-A-24
,, ')
Related Equipment X
/X A-25 Nonsafecy Loads on Class IX Power w/
Sources I
X A-26 Reactor Vessel Pressure Transient Protection (Overpressure)
X A-27 Reload Application Guide X
X A-28 Increase in Spent Fuel Storage Capacity X
X A-29 Design Fe atures to Control Sabotage X X
A-30 Adequacy of Safety-Related DC Power Supplies X
X A-31 RER Shutdown Require =ents X
X A-32 Evaluatien of Overall Effects of Missiles X
X A-33 NEPA Reviews of Accident Risks Environ = ental A-34 Instruments for Monitoring Radiati:n and ?rocess Variables During Accidents X X A-35 Adequacy of Offisite Pcwer Syste=s X X
A-36 Control of Heavy Loads Near Spent Fuel IX A-37 Turbine Missiles X
X A-38 Tornado Misciles X
X A-39 Determination of Safety Relief Valve (SRV) Fool Dynamic X
A-40 Seismic Design Criteria - Short Term1 X
Program AM himie Desien Criteria - Long Tor:f y
)
I CATEGORY B TECl!NICAL ACTIVITIES
- i RELEVANT TO
TASK NO.:
TITLE:
BWR PWR B-1 Environmental Tpel.nical Specifications X
X B-2 Forecasting Electricity Demand X
X B-3 Event Categorization X
~X B-4 ECCS Reliability-X X
l B-5 Ductility of Two-Way Slabs and Shells and Buckling Behavior of Steel Containments X
X l
li-6 Loads, Load Combinations, Stress Limita
,X X
g E
g B-7 Secontary Accilent Consequence Modeling X
l B-8 Locking Out of ECCS Power Operated Valves X
X B-9 Elect?lcal Cable Penetrations of Containment X
X l
1 B-10 Behavior of BWR Mark III Containment X
B-11 Subcompartment Standard Problems X
X B-12
. Containment Cooling Requirements (Non-LOCA)
X X
B-13 Harviken Test Data Evaluation X
,X B-14 Study of Ilydrogen H1xing* Capability in Containment Post-LOCA X X
B-15 CONTEMPT Computer Code Maintenance X
X B-16 Protection Against Postulated Pinping Failures in Fluid Systems Outelde Containment X
X B-17 Criteria for Safety-Related Operator Actions X
X 11 - 1 8 Vortex Suppression Requirements for C'ontainment Sumps X
11 - 1 9 Thermal-llydraulic Stability X
X h
I'
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X X
B-20 Standard Problem Analysis i
X X
j B-21 Core Physics i
B-22 LWR Fuel X
X j
LHFBRs B-23 LHFBR Fuel,
B-24 Seismic Qualification of Electrical and Hechanical X
X Equipment
{
B-25 Piping Benchmark Problems X
X B-26 Structural Integrity of Containment Penetrations X
X B-27 Implementation and Use of Subsection NF X
X B-28 Radionuclide/ Sediment Transport Program X
X y
B-29 Effectiveness of Ultimate Heat Sinks X
X em B-30 Design Basis Floods and Probability X
X B-31 Dam Failure Model X
X B-32 Ice Effects on Safety-Related Water Supplies Reactors in Northern f
United States I
B-33 Dose Assessment Hechodology X
X B-34 Occupational Radiation Exposure Reduction X
X B-35 Confirmation of Appendix I Ifodels for " Calculation of Releases of Radioactive Haterials in Gaseous and Liquid Effluents from Light-Water-Cooled Power Reactors" X
X B-36 Develop Design, Testing and Haintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption i
Units for Engineered Safety Feature Systems and for Normal Ventilation Systems X
X B-37 Chemical Discharges to Receiving Waters X
X B-38 Reconnaissance Level Investigations X
X D
~'
t,
~
~
l:
l B-39 Transmission Lines B-40 Effects of Power Plant Entrainment on P1,tnkton X
X B-41
' Impacts on Fisheries X
X B-42 Socioeconomic Environmental Impacts X
X B-43 Value of Aerial Photographs for Site Evaluation X
X B-44 Forecasts of Generating Costs of Coal and Nuclear Plants X
X B-45 Need for Power - Energy Conservation X
X B-46 Costs of Alternatives in Environmental Design X
X t
B-47 Inservice Inspection of Supports Class 1, 2, 3 and HC
.i Coponents X
X e
'd B-48 BWR Control Rod Drive Mechanical Failures X
B-49 Inservice Inspection Criteria and Corrosion Prevention Criteria for Containments X
X B-50 Post-Operating Basis Earthquake Inspection X
X B-51 Assessment of Inelastic Analysis Techniques for Equipment and Components X
X B-52 Fuel Assembly Seismic and LOCA Responses X
X L
B-53 Load Break Switch X
X B-54 Ice Condenser Containments X
B-55 Improved Reliability of Target-Rock Safoty-Relief Valves X
~
B-56 Diesel Reliability X,
X B-57 Station Blackout X
X B-58 Passive Hechanical Failures'.
X X
I l
~
5 B-59 N-1 Loop Operation in BWRs and PWRs X
X B-60 Loose Parts Monitoring Systems X
X B-61 Analytically Derised Allowable ECCS Equipment Out' age Periods X X
B-62 Re-Examination of Technical Bases for Establishing SLs, LSSs, and Reactor Protection System Trip Functions X'
X B-63 Isolatiori of Low Pressure Systems Connected to the Reactor l
Coolant Pressure Boundary '
X X
[i B-64 Decommissioning of Reactors X
X B-65 Iodine Spiking X
X Y
B-66 Control Room Infiltration Hessurements X
X e.
oo B-67 Effluent and Process Monitoring Instrumentation X
X B-68 Pump 0verspeed During a LOCA X
X B-69 ECCS Leakage Ex-Containment X
X l
B-70 Power Grid Frequency Degradation and Effect on Primary Coolant Pumps X
j B-71 Incident Response l
B-72 Health Effects and 1.ife-Shortening from Uranium and I
Coal Fuel Cycles X
Jt B-73 Honitoring for Excessive Vibration Inside the Reactor Pressure Vessel X
X a
e g
O s
~
G[
G..
~
~
.. - _. ~.
e
[
' CATECORY C TECilNICAL ACTIVITIES RELEVANT TO:
' TITLE:
4 l
C-1 Assurance of Continuous Long-Tern Integrity of Seals on Instrumentation and Electrical Equipment X
X C-2 Study of Containment Depressurization by Inadvertent spray Operation to Determine Adequacy of Containment X
X Exiernal Design Pressure C-3 Insulation Usage Within Containment X
X C-4 Statistical Hethods for ECCS Analysis X
X X
X f
C-5 Decay Heat Update f
X X
C-6 LOCA Heat Sources i
X l
C-7 PWR System Piping C-8 Main Steam Line Leakage Control Systems X
C-9 RHR Heat Exchanger Tube Failures X
X Lf C-10 Effective Operation of Containment Sprays in a LOCA X
X.
[$
C-11 Assessment of Failure and Reliability of Pumps and Valves X
X i
C-12 Primary System Vibration Assessment X
Non-Random Failures X
X C-13 C-14 Storm Surge Model for Coastal Sites X
X C-15 HUREG Report for Liquid Tank Failure Analysis X
X on coasts 1
C-16 Assessment of Agricultural Land in Relation to Power
~
Plant Siting and Cooling System Selection X
X farm land Interim Acceptance Criteria for Solidification Agents C-17 for Radioactive Solid Wastes X
X CATEGORY D TECIINICAL ACTIVITIES I
D-1 Advisability of a Seismic Scram X
X l
D-2 Faergency Core Cooling System capab()$ty,for future PlanFa X
X D-3 Control Rod prop Accident X
4
"The individual discussions of the safety im-portance of these tasks as they relate to the Black Fox Station are provided in a subsequent section of this testimony.
Although, based on our cursorv review of the preliminarv results of this risk-based evaluation, we can't exoect the final study to reveal that the staff has grossly erred in its initial judgments regarding tne safety significance or particular issues, it is possible that the evaluation will indicate that a reordering of the priorities of sume of the issues is appropriate.
In the eventuality that some reordering does take place, our specific plans and schedules for some of the Category A tasks discussed in this testimony may change in order to more effectively utilize resources.
In addition, some of the lower priority tasks could be elevated to a higher level of importance and accordingly work may begin on them at an earlier date than currently envisioned."
(emphasis added)
(26)
Finally, on October 6 tha NRC gave a presentation to the ACRS Full Committee meeting regarding the Staff's generic program progress.
Key points discussed were that the NRC is currently re-assessing the generic issues and will re-prioritize them in the near future.
The status of their review as of October 6 was described as " cursory" and being performed by use of " bounding approach" methods.
A copy of the visual aids used is attached to this report.
(Note page 3 which gives th latest prelMnary results of the highest impact items.)
Page
\\
4 lists moderate impact, but in the meeting the NRC indicated they would probably combine those two groupings.
They are in the process of writing a risk screening report which will develop into a firm list.
This report was expected to be issued in October of this year.
l 1
3-20
~
l s.
4 4.
REFERENCES 1.
NUREG-0410, NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants, January 1978, US NRC, Washington, DC, pages D-1 to D-7.
(Note:
NUREG-0371, entitled Aporoved Task Action Plans for Category A Generic Activities, Vol.
1, Revision 1, is contained as Appendix F of NUREG-0410.)
2.
Investigation'of Charges' Relatin'g'~to Nuclear Reactor Safety, Hearings Before the Joine Committee on Atomic Energy, Vol. 2,1976, U.S. Government Printing Office, Washington, DC, pages 1200 to 1445.
3.
Letter, M. Bender, ACRS Chairman to Joseph Hendrie, NRC Chairman, entitled, " Status of Generic Items Relating to Light-Water Reactors:
Report No.
6,"
dated November 15, 1977.
k-4.
Ibid 2, Vol. 1, page 47.
5.
NUREG-0138, Staff Discussion of Fifteen Technical Issu'es i
Listed in Attachment to Novemoer 3.
197o Memorancum from Director, NRC to NRC Staff, Novemoer 19 75, US NRC, t
Washington, DC.
6.
NUREG-0153, Staff Discussion of Twelve Additional Technical Issues Raised by Responses to Novemoer 3, 1976 Memorancum From Director, NRC to NRC Staff, Decemoer, 19 76, US N RC,
Washington, DC.
7.
Ibid 3.
8.
Ibid 3, page 2.
3
./
9.
Reginald H. Jones, " Impromptu Statement on Nuclear Energy",
to New Y'ork Society of Security Analysts, Hotel Pierre, New York City, December 17, 1975.
- 10. Report of the' Advisory Committee on Reactor Safeguards Work Group No. 4, March 25, 1976, page D-17.
- 11. Ibid 2, Vol. 1, page 187.
- 12. Ibid 2, Vol.1, page 187.
- 13. Ibid 2, Vol.1, page 883.
- 14. Letter, Joseph Hendrie, to Congressman John Dingell, February 9, 1978.
- 15. Ibid 14 3-21
- 16. Note that the Atomic Safety and Licensing Board issued a Board subpoena on October 15, 1978 to obtain the complete GE Reed Report in the Black Fox (Public Service Company of Oklahoma) construction permit proceeding.
- 17. Letter, Roger Mattson to Glenn Sherwood, July 10, 1978.
- 18. " Memorandum In Support of General Electric Motion To Quash", by GE, Black Fox Station, Docket Nos. 50-556 and U
50-557, October 30, 1978.
};
- 19. Ibid 2, Vol. 2, page '1203.
1 l
- 20. Ibid 1, page iii.
t i
- 21. Ibid 1, page 'iv.
- 22. Ibid 1, page 7.
- 23. Title 10, code of Federal Regu'lations, Part 50.109.
]
- 24. Ibid 1, Appendix B.
- 25. NUREG-0471, Generic Task Problem Descriptions, June,1978, US NRC, Washingcon, DC.
- 26. Testimony of Aycock, Crocker, and Thomas, Black Fox Station, September, 19781 5.
SUMMARY
AND CONCLUSION The Task Action Plans, GE Reed Report items, and ACRS generic j
items identify technical deficiencies that appear to be signi.' )
%)
cant to reactor design and operation.
The resulting safety j
].
uncertainty is one of the most serious factors affecting the assessment of the risk of the Shoreham plant.
Many of the unresolved issues mentioned in this report could quite easily change from " potential" to "real and urgent" through additional analysis of failures experienced in one or more operating plants.
knen problems are discovered which require correction after the plant goes into service and becomes radioactive, extremely 3-22
t long outages could be ' involved to bring the plant up to I
required safety standards.
Resolution of the generic problems i
may also affect equipment that is in or a part of the primary s
system.
These primary components become highly radioactive
,g due to deposition of activated corrosion products, or, as a
.[
result of direct irradiation of the material its&lf.
Repair work or modification to these components, therefore, becomes extremely costly due'to the difficulty in working with this material and to the large amount of manpower required to absorb the' radiation exposure that may be involved with the modifi-j e
1 1 "
cation.
Also, disposal of the removed parts must be handled as a controlled radioactive shipment and this in itself is hazardous.
The County, therefore, should assure themselves that LILCO and the NRC address these generic problems, and that resulting resolutions to the deficiencies are implemented in a timely manner.
'd 1
e 3-23
wa RES/NRR REASSESSMENT OF GENERIC ISSUES l
OBJECTIVE:
REVIEW CURRENT LIST.0F NRR'S. GENERIC ISSUES, AND ASSESS'THEIR RELATIVE IMPORTANCE' i
APPROACll:.ASSESSSdFETYSIGNIFICANCEBhARSS-Tve'E.SCREENINGPROCESS*
t FACTOR IN OTHER NON-RISK CRITERIA
~
RECOMMEND ADJUSTMENTS TO. ESTABLISHED PRIORITIES I
STATUS:
INITIAL SCREENING COMPLETED, REPORT BEING DEVELOPED g
9 is n
j 1-l t.
U
~
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wt en.
j
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r '..
~
REVIEW 0F GENERIC ISSUES - RISK APPROACH GEf1ERAL. APPROACH USE FINDINGS OF RSS & RSS - FOLLOW ON STUDIEC AS A BASE.0F DEPARTU,RE A.ND l
DETERMINE SIGNIFICANCE Oc GENERIC ISSUE BY ESTIMATING iTS RELATIVE CONTRIBUTION TO RISK.
USE BOUNDING hPPROACH TO ESTIMATE RISK CONTRIBUTION.
e y
TESTS 0F_ RLSK SLrdl!FICANCE F
~ '
RISK'0F EARLY AND LATENT FATALIT'IES RISK OF ACCIDENTAL l'0 DINE RELEASE l
POSSIBLE IMPACT ON ROUTINE RELEASES i
l POSSIBLE IMPACT ON OCCUPATIONAL ExeOSuRES 1
l )
=
PRELIMINARYRESULTSOFRISKREVIEW0FGENERICISSUES
.i TASK ACTION PLANS WHICH llAVE 6REATEST POTENTIAL IMPACT ON RISK AT!iS (A-9)
DYNAMIC LOADS IN BWR PRESSURE SUPPRESSION CONTAINMENTS (A-6, A-7, A-8, B-551 l
a SYSTEMS }NTERACTIONS (A-17)
~
SEISMIC DESIGN CRITERIA (A 110) l DESIGN FEATURES To CONTROL SABOTAG'E (A-29)
BWR N0ZZLE CRACKING (A-10)
I 6
STATION BLACK 0UT REQUIREMENTS (B-57):
ISOLATIONOFLOWPRESSuRESYSTEMSCONNEdTEDTORCPB.(B-63)
DESIGN BASIS FLOODS & PROBABILITY (B-30)
~
OCCUPATIONAL RADI ATI0li Exe0SuRE REDUCTION (B-3ll)
INSULATION USAGE WITillh CONTAINMENT (C-3)
~
Q g
. LJ q)
(;
PRELIMINARY RESULTS OF RISK REVIEW 0F GENERIC ISSUES TASK ACTION PLANS WilICH-HAVE MODERATE POTENTIAL IMPACT ON RISK S-GTuBEINTEGRITY(A-3SAII,'A-5)
NATER llAMMER (A-1) l FRACTURE TOUGHNESS OF S-G/RCP Supe 0RTS (A-12).
Y ASYMMETRIC BLOWDOWN LOADS (A-2) v.
DC Power SUPPLIES (d-30) l 4
i QUALIFICATIONOFC(ASSIESAFETYRELATEDEouiPMENT(A-2fi) 1 i
DECOMMISSIONING OF REACTORS (B-6ll) 1 s_.
r CROSS-COMPARIS0ll 0F GEllERIC TASKS H0sT SIGNIFICANT TAPS.-
ACRS "A'S" flRR. TAPS PER RISK SCREENING i
II-l TURBINE filSSILES A-37 11-3 PRESSURE VESSEL FAILURE BY A-ll l
l TilERMAL Sil0CK I
II-6 ilort-RANDOM MULTIPLE FAILUhtES A-9, A-30, A-35, I
A-9*, A-50
,i (TilREE SUBGROUPS)
B-56, B-57, C-13 11-7 Fuel BEllAVIOR UNDER ABNORMAL 3-22
~
i CoriDITIONS 0'
11-10 ECCS CAPABILITY FOR FUTURE (RES)
PLAllTS II A-3 S-G TUBE LEAKAGE A-3, A-4,.A-5 A-3,A-4,A-5 1
II C-2 DESIGN FEATURES TO CONTROL A-29
~
A-29" t
SADOTAGE l
II C-5 WATER llAMMER A-1 A-1 l
I Il C-7
~B'ilR f1 ARK } CONTAINNENr A-6,A-7,A-39 A-6*,A-7*
l II D-1B SYSTEMS INTEftACTIONS A-17 A-17 l -
O C)
~,
.s
s,
+
STATUS OF RE-EVALUATI0fl 0F GENERIC ISSUES PRELIMINARY SCREENING OF TAPS BY RISK SIGNIFICANCE.COMPi.ETED.
I
\\
TASK GROUP PREPARING DRAFT REPORT FOR NRR/RES/ACRS REVIEW.
~
(INCLUDES DISCUSSION OF APPROACH, OVERALL RECOMMENDATIONS / CON'LUSIONS)
C t
a O
0 a
O e
CONTENTION 4a SAFETY ISSUES RAISED BY NRC STAFF l.
STATEMENT OF CONTENTION 4a.
Intervenors contend that the Applicant and Regulatory Staff have not adequately considered individually a nu=ber of generic light water safety issues raised by NRC Staff members and applicable to Shoreham in accordar.ce with the backfitting requirements of 10 CFR Part 50.109 and/or the general design criteria of 10 CFR Part 50, Appendix A.
This contention includes, but is not limited to, the following design features for structures, systems, and ccmponents :
1.
Treatment of non-safety grade equipment in evaluation of postulated steamline break accidents.
ii.
Lack of independence on ECCS valves.
iii.
Analysis of postulated reactor coolant pump rotor seizure systems, iv.
Protection against single failures in reactivity control systems.
~
v.
Use of probalistic assessment of reliability.
vi.
Grid stability.
vii.
Interpretation of NRC General Design Criteria 19, "Contrcl Room."
viii.
Onsite and offsite emergency power.
ix.
Instrument trip setpoints in standard technical specifications.
x.
Automatic resetting of reactor trip system trip l
bistable relays.
xi.
Passive mechanical valve failures.
i xii.
Electrical cable penetrations of reactor contain-ment.
i xiii.
Analysis of the interaction of structures and the i
supporting soil.
4-1 i
e s
xiv.
Instruments for moitoring both radiation and process variables during accidents.
xv.
Safety i=plications of control system failures j
and plant dynamics.
\\
xvi.
Improving availability of offsite power.
xvii.
Improvement of BWR-shutdown reactivity performance.
2.
CONTENTION CHRONOLOGY This contention, along with all other currently active County contentions, was originally' described in County's Amended Petition to Intervene, dated September 16, 1977.
This contention was ruled accepta5 ale for purposes of discovery
^
i at the October 11, 1977 Prehearing Conference.
In the Board's ruling of January 27, 1978, companion Contention 4b was ruled to be unacceptable and the Board also repor*md that Contention 4a-rriii had been dropped by County.
The January 27, 1978 ruling also erroneously classified the uajority of Contention 4a as disallowed, even though it had been ruled acceptable for discovery at the Prehearing Conference.
The March 8, 1978 Board order subsequently verified that Contention 4a ite=s i through xvii were acceptable for discovery.
3.
BASIS'FOR CONTENTION The basis for this contention is as stated in the September 16, 1977 Amended Petition to Intervene', and as further elaborated upon in Councf's Response to Acplicant's Second Inte=ogatories dated January 31, 1978, pages 1 through 7.
In order to present 4-2
the current status of this centention, the following discussion of the basis for this contention includes information extracted I
from these two submittals.
The AEC in 1974 tabulated the unresolved safety issues into an, internal Technical Safety Activities Report (TSAR), but did not communicate to the public the requirements for resolution.
The TSAR in December, 1974 listed 223 items of concern to the NRC, of which 173 of the items were categorized as having "an important impact on the licensing review process."
- However, in 1976, following the resignation of Robert Pollard, the NRC apprently discontinued regular preparation and distribution o the TSAR.
By the fall of 1976, in response to a memorandum by' Rusche of the NRC, one or more members of the NRC Staff ide.nti-a fied 27 technical issues as " problems whose priority, progress, or resolution was, in their opinion, unsatisfactory."
The list of unresolved items identified by concerned engineers within the NRC is summarized in Table 4-1.
Contention 4a is based upon the adequacy of the resolution of those' items relevant to a B'JR for the Shoreham facility.
Applicable regulations for the listed safety issues are '
both 10 CFR 50.57 and 50.109.
50.57 is the primary section having to do with the issue of operating license.
The finding of " reasonable ' assurance---without endangering the health and l
safety of the public" is questioned in view of the fact that l
the Staff may not require the backfitting of the facility for issues to " provide substantial, additional protection which is l
required for the public health and safety" as required by 50.109.
l l
4-3
r The specific subsection of the Title 10 Code of Federal Regu-j lations and/or the particular criterion of 10 CFR Part 50, i
- [
Appendix A for each of the issues listed in Contention 4a, l,
i through xvii, are provided in the following tabular chart:
)
Safety 10 CFR Part 50 i
Issue Appendix A Criterion i
GDC 2 11 GDC 5, 35, 37 iii GDC 10 iv GDC 24, 25, 26, 29 v"
GDC 10, 20, 29
T vi GDC 17, 18 vii GDC 19, 20 i
viii GDC 17, 18 ix GDC 20, 21 x
GDC 13, 20 xi GDC 10, 12, 15 xii GDC 3, 4, 18 xiii GDC 2 xiv GDC 4,.13 xir GDC 13, 20 xvi GDC 17, 34, 35, 38 i :
xvii GDC 13, 20, 21 1
The NRC STaf'f members who raised the generic light water safety il issues identified in items i through xvii are fully identified in NUEG-0138, Staff Discussion of Fifteen Technical Issues 4
Listed in Attachment to November 3,1976 Memorandum from Director, NRR to NRR Staff and in NUREG-0153, Staff Discussion of Twelve Additional Technical Issues Raised by Responses to November 3,1975, Memorandum from Staff, NRR to URR Staff.
In addition, the ACRS reviewed the status of each of the' items 4-4
and issued a report summarizing their assessment.
The following is a brief discussion of some of the items listed in the con-i tention.
(i)
This safety issue does not apply to the Shoreham Plant being first identified as problem with a pressurized water reactor.
The concept of misapplication of non-safety grade equipment-does of course apply, but may be considered covered under other criteria and/or issues.
y (ii)
The question of lack of independence of ECCS valves was first raised by the Staff in the Jamesport and McGuire PWR plants.
The particu-lar valves in question for those two plants of course do not apply directly to Shoreham.
How-ever, Shoreham does have a substantial number of valves that are commonly used between ECCS, RHR, and similar systems.
Action taken by the Applicant, Staffandreactorequiptentsupplia]
to re-analyze the possibility of this problem existing on a new BWR line such as is applied at Shoreham should be identified to ensure that it has not been overlooked.
1 (iii)
This safety issue also was firs.t identified in in a ?WR plant review.
The concept of conce n identified in the response to Contention 1i l
4-5
above should be applied to assure the com-l P aceness of accident analyses regarding the l
Shorenam reactor coolant pumps.
(iv)
The safety issue identified in this item was first applicable to a PWR and involved concernsa over accidental rod withdrawal.
This same con-cern exists in essence at Shoreham for both the operating and shutdown condition.
The items of concern are the lack of design verification of the rod sequence control system used at Shoreham and the possibility of error leading to rod with-drawal errors.
For examples of these, see the event descriptionsof the accidental criticalities experienced at the Vermont Yankee UWR and at the Millstone BWR.
(v)
The concern of adequacy of assessment of relia-bility is well documented in NUREG-0138 and particularly applies to Shoreham, a relatively new model BWR.
The Staff's response to this item of concern on page 8-3 of this NUREG is based on the fact that anticipated transients without scram (ATWS) are not "the predominant contributor to core melt probabilitf for light water reactors (as calculated in WAS~d-1400)".
This statement is generally true, but it avoids the fact that transients are the predominant 4-6
contributor for BWR's.
This discrepancy i.s I
disturbing and may be indicatiV2 of a super-ficial evaluation of the applicability of this safety issue to the Shoreham BWR.
(vi)
The particular concern about Shoreham's " grid stability" is well described in NUREG-0138 and is further evidenced by problems with off-site power at Millstone 2, Turkey Point 3 and 4, and l
Indian Point 2 and 3.
Approved Task Action Plan i
e l
A-35, described in NUREG-0410, NRC Program fo.*
I l
the Resolution of Generic Issues Related to Nuclear Power Plants provides a good description l
of the problem, the concerns, and the proposed NRC plan.
(vii)
The concern over the capability of the Shoreham design to fulfill the 'equirements of GDC 19, r
Control Room, has to do with undemonstrated capa-bility of -he plant to independently and remo bring the reactor to a condition of safe, hot shutdown and subsequently to cold shutdown.
Adequacy of the design is given only a cursory description in the FSAR.
For additonal comments regarding this concern, please refer to the
?
(
October 7, 1977 ACRS meeting transcript in which Dr. Ebersole spoke of the " skillet syndrome" 4-7
s....
t (viii) See statement concerning Contention vi.
(ix)
No instrument trip set points have been identi-fied in the Technical Specification as the Technical Specification has not been issued in the FSAR as yet.
For description of concern, see NUREG-0138.
(x)
See NUREG-0153, Item 16.
l (xi)
Sea. generic Item B-58.
(xii)
See generic Item B-9.
l l
(xiii) See generic Item A-40 and A-41.
l (xiv)
The instruments in question are those that are well described in NUREG-0153, Sec ion 23 and are also those well described in REG Guide 1.97, the subject of mucli recent discussion.
This subject is also identified in NUREG-0410, and has an r.
approved Task Action Plan (Task A-34) in that document.
Plan of schedule for implementation l
of action on Shoreham is needed.
3 (xv)
See' generic item A-17.
l (xvi)
See generic Item A-35.
(xvii) Installation of an adequate Prompt Relief Trip 1
(PRT) system may mitigate this concern.
4-8 Q
e
4.
REFERENCES i
1.
Investigation of Charges Relating to Nuclear Reactor i
Safety, Hearings Before the Joint Committee on Atomic Energy, Vol. 2, 1976.
2.
NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants, (NUREG-0410).
3.
Generic Task Problem Descriptions - Category B, C, and D Tasks, (NUREG-04/1),,
4.
Staff Discussion of Fifteen Technical' Issues Listed in i
Attachment to Novemoer 3, 1976 Memorancum Irom DLrector, NRR to NRR Scatf,(NUREG-0135).
5.
Staff Discussion of Twelve Additional Technical Issues Raised 'oy Resoonses to Novem'oer 3, 197o Memorancum Irom j
Director, NRR to NRR Staff, (N UREG-013 3).
5.
SUMMARY
AND CONCLUSION I
For the foregoing reasons, it is clear that many of the safety issues identified in Contention 4a are still in a stata j
of dynamic consideration and have not been adequately considered f
by the Applicant and Regulatory Staff because of the incomplete-ness of review that exists to date.
Following issuance of the
(
SER, the. status of each of these items should be re-reviewed.
The re-review should also address any updates of NUREG-0410 an]
l NUREG-0471.
l I
i f
r l
1 i
1 i
4-9 1
o....
CONTENTION Sa, b, & c QUALITY ASSURANCE / QUALITY CONTROL
+
1.
STATEMENT OF CONTENTION Sa.
Intervenors centend that the Applicant and Regulatory Staff have not adequately demonstrated that the quality assurance program for the design and installation of struc-tures, systems, and components for the Shoreham nuclear station was conducted in a timely manner in compliance with the pertinent portions of 10 CFR Part 50, Appendix B, Cri-teria 1 through 18 and with 10 CFR Part 50, Appendix A, Criterion 1 with regard to:
1.
Equipment qualification and the resulting qualification records.
s 11.
Organizational independence of inspection and construction supervision.
iii.
Applicant's quality program not implemented until after a large portion of plant struc-tures, systems, and components are designed and inscalled.
iv.
Control of clean rooms and other special environmental conditions.
v.
Training, qualification, and _ assessing of welders and welding procedures.
vi.
Internal audit frequency and reporting of results to management.
vii.
Traceability of welding records to welders.
viii.
Docunentation of test results.
ix.
Calibration and accuracy of measuring and test equipment.
x.
Qualification and training of NDT personnel.
xi.
Acceptance of incocing equipment including st.itable status indication, such as tags or labels.
xii.
Release of uninspected equipment for installation.
i xiii.
Indentification, segregation, review, and release of nonecnfor=ing items.
5-1 l
e 3
xiv.
Records of nonconforming items.
)
i xv.
Test records.
?!
l xvi.
Chronological records of detection and resolution
- j of all safety-related problems detected during i
construction.
l l
xvii.
Revision control of design documents.
.i xviii. Selection, evaluation, and source inspection for j
suppliers.
I xix.
Technical adequacy of procurement documents.
1 xx.
Control of changes to procurements documents.
i Sb.
Intervenors contend that the Regulatory Staff's Inspec-tion and Enforcement (IE) Program has not adequately verified that the Applicant's quality assurance program for Shoreham has been implemented in accordance with the requirements of
.)
l 10 CFR Part 50.34(a), paragraph 7 and the pertinent portions of 10 CFR.Part 50, Appendix B, Criteria 1 through 18 in that:
i 1.
The IE Program has identified only the symptoms of the Shoreham quality deficiencies, and has not
' required Applicant to initiate corrective' action to resolve the root causes, ii.
The IE's reliance on the Applicant for primary inspections at Shoreham with NRC officials serving as auditors has recently proven to be inadecuate in timely identifying quality deficiencies de other nuclear facilities.
(e.g., Browns Ferry, North Anna, Davis Besse, and Racho Seco.)
111.
The IE Program has no objective and quantitative
]
measurement of the Shoreham quality program effec-
./
tiveness.
Sc.
Intervenors further contend that the quality asurance program description for operation of the Shoreham plant, as provided in the FSAR, does not comply with 10 CFR Part 50.34 l
(b)(611), and with the pertinent portions of 10 CFR Part 50, Appendix B, Criteria 1 through 18 with regard to:
1.
Failure to address, as a minimum, each of the cri-teria in Appendix B in sufficiene detail to enable an independent reviewer to deter =ine whether and how all the requirements of the Appendix will be satisfied.
l l
5-2 l
ii.
Failing to adequately describe the extent ed which the operations phase quality assurance program will follow the guidance in WASH-1284, WASH-1283, and WASH-1309.
iii.
Failing to adequately identify, report, and analy=e all equipment failures discovered during operation and maintenance.
iv.
Failing to ensure that replacement materials and parts are equivalent to the original equipment, that replacements are installed in accordance with adequate process procedures, and that the repaired or rewored structure, system, or com-ponent is adequately inspected and tested.
2..
CONTENTION CHRONOLOGY This contention, along with all other currently active
\\
County contentions, was originally described in County's Amended Petition to Intervene, dated September 16, 1977.
All parts of this contention with the exception Sc(ii) were ruled as acceptable for discovery at the October 11, 1977 Prehearing Conference.
The continued acceptability of all such parts was confirmed in the Board's order of January 27, 1978 and in addition, Part Sc(ii) was included as acceptable for discovery
. subject to the condition that the County would be required to
( '/
file a precise specific contention for Sc(ii) at the cloy.e of discovery.
3.
BASIS FOR CONTENTION The basis for this contention is as stated in the A= ended Petition To Incerview dated September 16, 1977, and as further elaborated upon in the County's Response to Applicant's Second r
Interrogatories dated January 31, 1978, pages 3 through 14 In order to present a co=plete status of the contention in this report, the following discussion includes extracts from 5-3
the preceding two documents along with relevant new infor-mation.
l Appendix B, '.' Quality Assurance' Criteria For Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50 requires, in part, that "every applicant for an opera-tive license is required by the provisions of Part 50.34 to include in its Final Safety' Analysis Report (FSAR) a de-scription of the quality assurance program for operations.
The 18 criteria of Appendix B set forth the quality. assurance requirements which " apply to all activities affecting the safety-related functions of those structures, systems, and components. "
In addition, General Design Criterion'1,
" Quality Standards and Records," of Appendix A to 10 CFR Part 50 describe quality assurance requirements.
The latest regulatory requirements are, in addition, de-fined by the Staff in the following Regulatory Guides and documents:
a.
Regulatory Guide 1.28, March 1978 (ANSI N45.2-1977) b.
Regulatory Guide 1.30, August 1972 (ANSI N45.2.4-1972) c.
Regulatory Guide 1.37, March 1973 (NASI N45.2.1-1973) t d.
Regulatory Guide 1.38, May 1977 (ANSI N45.2.2-1972)
Regulatory Guide 1.39, September 1977 (ANSI N45.2.3-1973) e.
f.
Regulatory Guide 1.58, August 1973 (ANSI N45.2.6-1973) i g.
Regulatory Guide 1.64, June 1976 (ANSI N45.2.11-1974) l l
h.
Regulatory Guide 1.74, February 1974 (ANSI N45.2.10-1973) l l
5-4
i.
Regulatory Guide 1.88, October 1976 (ANSI N45.2.9-1974) j.
Regulatory Guide 1.97, ' April 1976 (ANSI N45.2.5-1974) k.
Regulatory Guide 1.116, liay 1977 (ANSI N45.2.8-1975) l '.
Regulatory Guide 1.123, July 1977 (ANSI N45.2.13-1975) m.
ANSI N45.2.12-1977 The Applicant's description of quality assurance measures are summarized in Section 17 of the FSAR.
3a.
Implementation of Quality Program During Construction As stated in County's Amended Petition to Intervene, Page 6, 10CFR 50 Appendix A, Criterion 1 applies to all twenty of the issues (i through xx) listed in Contention 5 (a).
The pur-pose of discovery, of course is to review the QA records for design, manufacture, and construction in order to further particularize the precise details of~the issues and in-adequacies identified in the contention.
County's Request No. 1 for the Production of Documents dated December 23, 1977 requested that the Applicant produce docu=ents for the County I
s,/
to assist in the determination requested in this question.
l Applicant's January 17, 1978 response to this request indi-cated that four listed Lilco documents would be made avail-able for County's review during normal business hours.at Lilco 's corporate headquarters.
County's consultants con-ducted an initial review of these docu=ents on January 30, 1978.
Applicant's respense for the production of other documents for County's use was a refusal to produce claiming they contained confidential commercial information and also claiming the requests were too broad in scope.
Later the Applicana did 5-5
agree to provide the County with an index of the requested quality records.
In order to re, view the detailed records i
(such as radiographs)a proprietary agreement will be re-quired to permit County and County's consultants access to the QA records.
Such an independent audit of QA records should be conducted by the County.
In addition, the County should review the NRC I & E reports related to Shoreham, as well as the I & E reports describing the QA inspections at the (1) (2) major vendors to Lilco (such as G.E.).
333,p
,3 of y
recent I & E reports documenting Lilco's failure to implement the required QA measures for weld rod control and equipment storage are included as Attachments A and B of this report.
It is also possible at this time to identify.the specific criteria of 10 CFR 50 Appendix B which do apply to the twenty issues identified in Contention 5 (a).
The. specific criteria are as follows:
Contention 5 (a) Issue Accendix B Criterion i
III, XI, XVII 11 I, I iii II iv XIII l
l v
II, II, IVII vi XVIII vii VIII, IX 1'
viii V, XI, XIV, IVII ix XII x
XVI, XI, XIV, XVII 5-6
a.
P xi VII, XIV xii VII, XV xiii XIII, XV xiv XV, XVII x
XVI, XI, XIV, XVII xv XIV, XV xvi XIV, XV xvii III, V, VI xviii IV, VII xix III, IV
(
xx IV 3b:
NRC I & E Program Effectiveness The NRC's Inspection and Enforcement: (I&E) program is c
intended to provide an independent verification that the Shoreham structures, systems, and components are designed, manufactured, installed, and operated Jin strict accordance with the applicable quality, assurance requirements.
In the past the I&E program has not fulfilled this intended function.
A recently released study conducted by the General Accounting Office (GAO) described the following weaknesses in the I&E i
program during nuclear power plant construction:
"Although the Nuclear Regulatory Commission is re-
{
sponsible for assuring that nuclear pcwer plants are constructed safely, it has not been independently testing the quality of construction work.
The Commission should do this, plus
--improve its inspection and reporting practices,
--use the inspectors' time and talents mere j
efficiently, and I
S 'T we~+vs,--+.
--better document its inspection findings.
j The Commission is aware of 'the need for improvements and has made some changes, one of which is the e
assignment of resident inspectg*p to selected reactors under construction." U>
i To evaluate the manner in which NRC inspectors did their work, GAO reviewed inspection reports at all five NRC regional offices and selected individual report 5.tems for detailed review.
During visits to 6 of the 7 powerplant construction sites, GAO reviewed 45 of the NRC inspection report items to determine if they could retrace the steps of the NRC in-
-]
i spectors, identify and review the documents the Staff re-viewed, and interview the site personnel the Staff contacted.
In some cases GAO could not determine which records the NRC inspector had reviewed because that infor=ation was not shown in the reports.
In other cases, GAO found errors in the reports which made it difficult for them to follow the work performed by NRC.
In addition, GAO ' identified instances where NRC inspectors ' overlooked or did not report certain weaknesses which GAO believe they should have found and re-5 ported.
Based upon the GAO findings and NRC's responses, GAO con-cluded that 31 of the 45 inspection report items, about 69 per-cent, were deficient in some manner.
GAO noted, however, that some of these items are insignificant and others re-flect their judgment as opposed to NRC's.
Also, while GAO did not attempt to determine the safety significance of these 5-8 S
=
t inspection deficiencies, NRC did not consider any of them major safety concerns or items of noncompliance with re-gulatory requirements.
The following chart su=marizes the results of the GAO detailed review of selected NRC report items, including the review of Shoreham:
NRC No. items Items deficient region Powerplant reviewed Ncmcer Percent I
Salem 5
5 Shoreham 10 8
~
Subtotal Region 1 15 13 87 II Sequoyah 6
6 100 III La Salle 8
6
.75 IV Arkansas Number One 9
1 11 V
Diablo Canyon
-7
-5 71 TOTAL 45 31 69
he nature and number of each type of deficiency noted for each NRC region are shown below!
Nature of Regions deficiency I
II III IV V
Total
, Inadequate reporting 7
3 1
0 1
12 Inadequate attention to l
details 1
3 2
1 3
10 Acceptance of inadequate li-censee action 3
0 1
0 0
4 Inadequate in-vestigation 2
0 2
0 1
5 TOTAL 13 6
6 1
5 31 l
5-9
d Other studies conducted by and for the Staff have j
identified deficiencies in the.c,onduct of the I&E program.
4 5
In particular, questions need to be answered about the Staff policy of relying on builders for primary inspec-tions with NRC officials serving as only auditors.
Because of the history of past I&E failures and the many proposed changes to I&E practices, the NRC should clearly describe in the SER the extent of the proposed I&E program for Shoreham.
Since a key factor in assessing the potential risk of a nuclear plant is the assumption of a disciplined, thorough quality assurance program.
Any inadequacies in the" Shoreham operating licensing review of the quality program may allow deficiencies in the program implementation which will pose a significant hazard to the public health and safety.
3c.
QA Program for Ocerations NRC document WASE-1309, " Guidance en Quality Assurance Requirements During the Construction Phase of Nuclear Power Plants," WASH-1283, " Guidance on Quality Assurance Requirements During Design and Procurement Phase of Nuclear Power Plants,"'-
and WASE-1284, " Guidance in Quality Assurance Requirements During the Operations Phase of Nuclear Pcwer Plants" contain ANSI quality assurance standards which describe requirements for implementing various portions of the quality assurance program.
A number of the standards contained in WASH-1309,
~
WASH-1283, and WASH-1284, were in draf t form.
Thus, in a 5-10
e number of instances, the provision of the WASH documents applicable to Shoreham have been* superceded by the issued ANSI standards and the accompanying Regulatory Guides.
The Staff procedures for reviewing compliance to WASH-1309, WASH-1283, and WASH-1284 are summarized as follows in Standard Format and Contents of Safety Analysic Reoorts for Nuclear Power Plants (PB-245-724, Revision 2, October 1975):
"Where a portion of the QA program to be implemented will follow the guidance provided by a regulatory guide, WASH-1283, WASH-1309, or WASH-1284, the program desc' rip-tion may consist of a statement that the guidance will be followed for that portion of the QA program.
When these documents are used in describing the QA program, the applicant should indicate how the guidance documents will be acclied to cortions of the OA crogram and should delineate the organizational element responsible fo'r implementing various provisions of the respective guidance documents within each major organization in the proj ect, including that of the applicant, the architect-engineer, the nuclear steam system supplier, the constructor, the construction manager (if other than the constructor)."
(Emphasis added)
The Applicant in the FSAR, ' Appendix 3B, indicates the in-tent to conform to some of the NRC Regulatory Guides which
[
address quality assurance.
The Applicant's agreement to in-plement the current Regulatory Guides is clearly not evident from the FSAR.
Furthermore, on page 17-2 of the S'tandard Format and Content of SARS (PB-245 724) the NRC states that "where a portion of the QA Program to be implemented will follow the guidance provided by a regulatory guide.....the Applicant should indicate how the guidance document will be applied to portions of the QA program."
(Emphasis added)
The Applicant's brief and general description of the Shoreham 5-11
1 QA program in the FSAR does not describe how the relevant l
regulatory guides on QA will be. implemented.
The procedures which describe how the Shoreham QA program is to be accomplished may be described in the Quality Assurance a
i; i
Manual and its implementing procedures rather than in the l
FSAR.
For power plant owners with little or no previous
?
nuclear power plant experience, such as Lilco, it is in-perative that the how of the QA program be clearly defined and described in the PSAR, as required by 10 CFR, Part 50.34, which states that "the description of the quality assurance
.}
program.....shall include a discussion of how the applicable requirements of Appendix B will be satisfied."
(Emphasis added)
Similar statements concerning the necessity to describe how the program will be implemented are included in the Section 17.1 of the Standard Review Plan, (NUREG-75/OS7) and page 17-1 of the' Standard Format and Content of Safety Analysis Report for Nuclear Power Plants (PB-245-724).
- Clearly, a..
i' cursory recapitulation of the 18 criteria of Appendix 3 in
- ]-
t the FSAR does not meet the requirements of 10 CFR, Part 50.34.
4.
REFERENCES l.
NRC I&E Reports for Shoreham Vendor Inspection Vendor Inspection Program 2.
NRC I&E Reports (General Electric design and/or manufacturing plants particularly).
3.
EMD-78-80, "The Nuclear Regulatory Commission Needs to Agressively Monitor and Independently Evaluate Nuclear Power Plant Construction," GAO, September 7, 1978.
5-12 l
4.
Examples include:
(a)
NUREG-0321, i studv>of the Nuclear Regulatory Commission Quality Assurance Prozram, August 19 7 7, U. S. N RC,
(b)
NUREG-0397, Revised Insoection'Prozram for Nuclear Power Plants, March 1978, U. S. NRC (c)
NUREG-0425, NRC Insoection Alternatives, February 197.
U. S. NRC.
(d)
GAO Report EMD077-30, Allegations'of Poor Con-struction Practices on the North Anna Power-plants, June 2, 1977.
(e)
Browns Ferry Nuclear Plant Fire, Hearings before the Joint Committee on Atomic Energy, September 16, 1975.
s (f)
NRC Press Release 76-122, Independent Assessment of NRC Quality Assurance Activities Planned, May 25, 1976.
t J
, y #
5-13
5.
SUM 4ARY AHD CONCLUSIONS The FSAR and the cited reports critiquing the NRC I&E program provide an inadequate basis for the County to con-clude that the quality assurance program of the Applicant and the NRC to be provided for the safety related structures,
systems, and components of the Shoreham Nuclear Power Stations, provides a reasonable assurance of safety.
For all the fore-going reasons, the County should request the Applicant to amend the FSAR to reflect the recommended analyses, review, and documentation of the quality assurance program as de-scribed in this report.
Any resulting modifications to Category 1 structures, systems, and components should be defined prior to the issuance of an operating license for Shoreham.
i I
9 e
i l
l l
5-14
.L.
.g'" "4e.
unirso synes ATTACHMENT A
. f.,,
NUCLEAR REGut.ATORY COMMisstON
$ ' h.' f..i RsGnoN e
, E.- V su n a vereve S, A.b 5
waac or anussia eencesvsvania iS*os
%....c s
Docket No. 50-322 M3 M3 ggd d
/
,.Long Island Lighting Company ATTN: Mr. Andrew W. Wofford h (,h_/)1. LCM Ipj tdCf Vice President 175 East Old Country Road Hicksville, New York 11801 Gentlemen:
Subject:
Inspection 50-322/78-02 This refers to the inspection conducted by fir. A. Toth of this office en February 15-17, 1978, at the Shoreham Nuclear Pcwer Station, Shcreham, x,
New York, of activities authorized by NRC License fio. CPPR-95, and to the discussions of our findings held by Mr. Toth with Messrs. F. Gerecke and T. Burke of your staff at the conclusion of the inspection.
Areas examined during this inspection are described in the Office of Inspection and Enforcement Inspection Report which is enclosed with this letter. Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observaticns by the inspector.
Based on the results of this inspection, it appears tha-t one of your
,r activities was not conducted in full compliance with NRC requirements, as set forth in the Notice of Violation, enclosed herewith as Appendix A.
This item of noncompliance has been categorized into the levels as described in our correspondence to you dated December 31, 1974.
This notice is ser.t to you pursuant to tne provisions of Section 2.201 of the (j
NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulaticns.
Section 2.201 requires you to submit to this office, within thirty (30) days of your receipt of this notice, a written statement or explanation in reply including:
(1) corrective steps which have been taken by you and the results achieved; (2) corrective steps which will be taken to avoid further items of noncompliance; and (3) the date when full cc=pli-ance will be achieved.-
In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosures will be placed in the NRC's Public Occument Recm.
If this A-1
~
Long Island Lighting Company 2
h report contains any information that you (or your contractor) believe to be proprietary, it is necessary that you make a written application within 20 days to this office to withhold such information from public disclosure. Any such application must te acccmpanied by an affidavit 4
executed by the owner of the information, which identifies the document 1
br part sought to be withheld, and which contains a statement of reasons l{
which addresses with specificity the items which will be considered by the Commission as listed in subparagraph (b) (4) of Section 2.790. The information sought to be withheld shall be incorporated as far as possible into a separate part of the affidavit.
If we do not hear frem you in this regard within the specified period, the report will be placed in j
the Public Document Room.
l Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely,
}
',& K W Robert T. Carlson, Chief Reactor Corstruction and Engineering Support Branch
Enclosures:
1.
Appendix A, Notice of Violation 2.
Office of Inspection and Enforcement
/
Inspection Report ?! umber 50-322/78-02 cc w/encis:
Themas J. Burke, Project !!anager Edward M. Barrett, Esquire Edward J. Walsh, Esquire
=
T. F. Gerecke,tianager, Engineering QA Department bec w/encls:
IE Maif & Files (For Appropriate Distribution)
Central Files Public Dccument Rocm (PCR) local Public Occument Rocm (LPOR)
Nuclear Safety Informaticn Center (i! SIC)
Tcchnical Infomation Center (TIC)
REG:I Readina Rcca State of !!ew' York A-2
l-
~
~
.)
i f
License No. CPPR-95 APPEi! DIX A NOTICE OF VIOLATION Based on the results of the NRC inspection conducted on Feb'ruary 15-17, 1978, it appears that one of your activities was not conducted in full compliance with conditions of your NRC Facility License No. CPPR-95 as indicated belcw.
This item is an infraction.
10 CFR 50, Appendix B, Criterion V requires, in part, that "Activi-ties affecting quality shall be prescribed by documented instruc-i tiens, procedures,...and shall be acccmplished in accordance with i
these instructions, procedures...." The Shoreham Nuclear Power i
Station FSAR Section 17.1.5A requires that suppliers of safety
("
related materials and services are responsible for imposing the above requirements on their internal operations.
i Project specification SH1-258B is imposed upon the various con-i tractors working in the containment drywell through their QA Manuals l
or contract documents. Specification SH1-2588 (dated December 1, i
1975) Part 2.1.5 requires that "All welding filler metal and the materials which have been issued and then not used during a given l
~
work shift shall be returned to the control area." Part 2.1.b requires that "All materials which are damaged shall be scrappel."
Contrary to the above, on February 15, 1978, twenty unused and
/
unreturned low-hydrogen type E7018 weld electrodes of varicus diameters were fcund in scattered locatiens inside the contaimaent drywell where safety related welding work activities were in progress.
This was in addition to significantly greater quantities of partly
?
used electrodes also lying loose throughout these work areas.
t F
i e
A-3 re
.._._ _;c Long Island Lighting Company 2
DEC 2 31577 i
In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosures will be placed in the NRC's Public Document Room.
If this report contains any information that you (or your contractor) believe tc be proprietary, it is necessary that you make a written application 2,
within 20 days to this office to withhold such information frem public disclosure. Any such application must be accompanied by an affidavit executed by the owner of the information, which identifies the document or part sought to be withheld, and which contains a statement of reasons which addresses with specificity the items which will be considered by the Ccomission as listed in subparagraph (b) (4) of Section 2.790: The inforuation sought to be withheld shall be incorporated as far as possib into a separate part of the affidavit.
If we do not hear frem you in this regard within the specified period, the report will be placed in the Public Occument Rocm.
Should you have any questions concerning this inspection, we will
)
pleased to discuss them with you.
Sincerely, f
[E y Robert T. Car son, ChiefReactor C Support Branch
/
e__i__..___.
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a 4
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e 3-2 e
9 1
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ATTACHMENT B
- t "' C%
unirro mit.s
,k,
NUCt. EAR nEGULA TONY r.u!.;.11ssloN f '4. :d,;"
ns.csoN I
- 2. M.y ;p 5
u, er.an mewa s ; *..;,1,/f o4 x No op enussia, asunsvi.vaniA iseos Docket No. 50-322 DEC 2 319n '
l Long Island Lighting Company ATTN: Mr. Andrew W. Wofford Vice President 175 East Old Country Road Hicksville, New York 11301 Gentlemen:
Subject:
Inspection 50-322/77-23 e
This refers to the inspection conducted by Mr. A. Toth of this office on s
December 5-9, 1977, at the Shoreham Nuclear Power Station, Shoreham, New :
York, of activities authorized by NRC License No. CPPR-95, and to the discussions of our findings held by Mr. Toth with Mr. W. Uhl of your staff at the conclusion of the inspection.
Areas examined during this inspection are described in the Office of Inspection and Enforcement Inspection Report which is enclosed with this '
letter.
Within these areas, the inspection consisted of selective examinations of procedures and repres.entative records, interviews with personnel, measurements made by the inspe.ctor, and observations by the inspector.
/
1 Based on the results of this inspection, it appears that certain of your l
activities were not conducted in full ccmpliance with NRC requirements, as set forth in the Notice of Violation, enclosed herewith as Apcendix
(_).
A.
These items of noncompliance have been categorized into the levels as described in our correspondence to you dated December 31, 1974. This notice is sent to you pursuant to tha provisions of Section 2.201 of the 4
NRC's " Rules of Practice," Part 2, Title 1.0, Code of Federal Regulations.
Section 2.201 requires you to submit to this office, within thirty (30) days of your receipt of this notice, a written statement or explanation in reply including:
(1) corrective steps which have been taken by yo; and the results achieved; (2) corrective steps which will be taken to avoid further items of noncompliance; and (3) the date when full compli-ance will be achieved.
I i
i 2
31 l
i
License' No. CPPR-95 APPENDIX A 1
NOTICE OF VIOLATION Based on the results of the NRC inspection canducted on December 5-9, 1977, it appears that certain of your activities were not cond cted in full compliance with conditions of your NRC Facility License No. CPPR-95 as indicated below. "These items are infractions.
A.
10 CFR 50, Appendix 8, Criterion V, states, in part, that: "Activitie affecting quality shall be... accomplished in accordance with (these) instructions, procedur!s or drawings."
The SNPS-FSAR, Chapter 17.1.5A states, in part, that: "The LILCO EQA Program establishes provisions for assuring that activities which affect the quality of safety related... components...are performed...in acccedance with... instructions, procedures and drawings."
'}
Project Procedure 10, revised August 1,1975, prescribes level "B" storage for non-manual valves which includes protection from weather.
Contrary to the above, on December 6,1977, two safety related, r.on-manual valves, numbers T46-TCV024A and T46-TCV025A, were un-covered and exposed to the rain in the short term storage area adjacent to the reactor building.
B.
10 CFR 50, Appendix B, Criterion XII states, in part, " Measures j
shall be established to control... preservation of materials....
When necessary, for particular products, special protective environ-ments...shall be specified and provided."
Q9ality Control Procedure 17.1 revis' ion B, date Section 3.4.1 entitled " Level A",
states in part, "This level i for items which require...an atmosphere free from dust...."
Contrary to the above on December 7,1977, the licensee level storage A did not conform with the requirements of the Quality Control Procedure 17.1 in that an atnosphere free from dust was not maintained as evidenced by significant dust on the shelves and safety related items in storage.
g 3-3 e
9
CONTENTION 6a & b INCLUSION OF ALL CREDIBLE LARGE ACCIDENTS IN SAFETY EVALUATION AND DESIGN BASIS 1.
STATEMENT OF CONTENTION 6a.
Intervenor, contend that the Applicant and Regulatory Staff have not adequately demonstrated that all credible or reasonably possible accident mechanisms have been considered and included in the design base as required by 10 CFR Part 50, Appendix A, Criterion 14, 31, and 35, with regard to the following:
1.
Experience and susceptibility of BWR pressure vessels to failure due to cracking at radius blend of feedwater noznles.
ii.
Possibility of failure of feedwater nozzle failure at location between pressure vessel and biological shield causing gross movement of vessel and subse-quent failure of reactor internals and/or ECCS.
capability.
iii.
Possibility of gross vessel movement resulting from reactivity addition accidents initiated by failure of inadequately designed system for control of rod sequence or other single failure induced transient, resulting in gross vessel internal failures and/or loss of ECCS function.
iv.
Gross vessel movement and/or damage causing loss of ECCS function as a result of sabotage or intentional misoperation.
-,/
6b.
Intervenors further contend that the Applicant and Regu-latory Staff have not adequately demonstrated compliance with 10 CFR 100.11, Determination of Exclusion Area, etc., as a result of not considering as credible events, accidents such as those listed in "a" above.
2.
CONTENTION CHRONOLOGY This contention, along with all other currently active County contentions, was originally described in County's l
Amended Petition to Intervene, dated September 16, 1977.
This 6-1
contention was found acceptable by the Board at the Prehearing I
Conference held October 11, 1977.
3.
BASIS FOR CONTENTION The basis for this concention is as stated in the September 16, 1977 Amended Petition to Intervene, and as further expanded upon in County's Response to Applicant's Second Interrogatories dated January 31, 1978, pages 17 through 19.
In order to present a complete s tatus in this report, the i
following discussion of basis for contention has been extracted from those two documents.
The basis for contention 6a is 10 CFR Part 50, Appendix A, Criterion 14, 31, and 35.
The possibility of other large accidents sequences is further affected by the lack of adequate demonstration of the rod sequence control system (RSCS)
The regulatory basis for the RSCS system is contained in 10 CFR Part 50, Appendix A, Criterion 15 and~29.
A description of the RSCS is found in the FSAR on page 15.1-32b.
This description is simplistic and cursory in nature and states only that the RSCS " prevents the selection of an out of sequence control rod" and concludes that there is no basis for rod withdrawal error in the start up power range.
The RSCS was an added on element in the Shoreham con-trol system design and has never been fully qualified by testing with all interconnecting systems and under the conditions of ecmplete in':eraction of the control system.
Further, WASH-1400 6-2 I
e
- a t
indicates that transients are a major contributor to the i
initiation of accidents in the BWR.
This finding makes it doubly important that the verification and adequacy of systems at Shoreham designed to prevent such transient initiated acci-dents be suitably demonstrated.
Examples of other transients of concern in this category are rod ejection, ganged rod with-drawal, rod drop accident, or other accidents which could result-8-
)
in the rapid addition of reactivity.
l Evidence that the " Applicant and Regulatory Staff have 4
l
' not adequately demonstrated that all credible or reasonably i
possible accident mechanisms have been evaluated and included
{
s in the Shoreham design base" is indicated by the following:
1 i
1)
The susceptability of BWR pressure vessel failure i
to cracking at the radius blend of feed water nozzles is well known, current, and well documented.
Two significant repoi ts are NUREG-0312 and General Electric s
Report NEDO-21480.
Additional information on this i
problem is contained in NUREG-0410, NRC Program for l;
the Resolution of Generic Issues.
Feed water no::le
- s. /
j-cracking has been designated a "C~ategory A" problem
]
and has been assigned approved Task Action Plan 'A-10.
This plan calls for final guidance to be issued to Applicants in October, 1979, so it is obvious that the review applicable to Shoraham is not complete at this time.
2)
The subject of assymetric blowdown loads on the eactor vessel is also discussed in NUREG-0410.
It is listed as a Category A item and Task Action Plan 6-3
_7_
A-2 has been approved. ~ The ' discussion in this document is primarily centered on PWR's, so a l
cursory review of this document might indicate it does not apply to Shoreham.
However, this sub-ject as it applies to BWR's was discussed in detail at the 'ACRS Fluid / Hydraulic Dynamic Effects Sub-committee meeting in San Francisco on November 30, 1977.
At that meeting, a presentation was made by General Electric addressing their review of assymetric blow down loads on the most current design BWR-6.
When asked as to the applicability of the analysis to r']
earlier BWR models (such as that being constructed at Shoreham), it was admitted that it did not ' apply.
Adequacy has, therefore, not been demonstrated as reviewed during the discovery process.
3)
Consideration should be given to the possibility of
~
l
" Class 9 Accidents" being a required analysis for j,
Shoreham.
Precedent for this may emerge through Class i
9 analysis that is under consideration for the off- ;].
shore power plant (barge) liquid pathways study.
Sub-stantial reason exists for this to be also required for Shoreham because of its close proximity to the shore and because of the unique water table character-istics existing at the site.
The fact that indecision exists within the regulato y l
agency as to the need for further " Class 9 Analyses" is 6-4
i e
I evidenced by the fact that the March 7, 1978 letter from the NRC's Mr. Case to the Commissioners (SECY-78-137) recommends the inclusion of core melt considerations in site comparisons.
Contention 6b addresses the determination of the plant exclusion area as required by 10 CFR 100.ll(a).
The basis for this contention is the Note 1 therein which states that:
"The fission product release assumed for these calculations should be based rpon a major acci-dent, hypotheisized for purposes of site analysis or postulated from considerations of possible
~
accidental events, that would result in potential i
hazards not exceeded by those from any accident l
considered credible."
(emphasis addea)
(
Applicant's Second Interrogatories dated December 8,1977 requested additional information on Intervenors' release assu=p-1 tions and the Board's March 2,1978 order compelled such infor-
{
mation (Question F.1, Contention 6b).
As indicated above, there is reason to believe that major t
i accidents involving gross core disruption, vessel movement, or vessel rupture are credible and should be considered.
These
^
events would result in fission product release categories similar
'w '
to the BWR-1, 2, and 3 events described in WASH-1400, and appro-priate analyses should be performed.
l
- 4. RRFERENCES I
l.
The Reactor Safety Study (WASH-1400)
(NURE, -75 / 014).
G 2.
Risk Assess =ent Review Group Report, USNRC (NUREG/CR-0400; 3.
NRC Program For Resolution of Generic Issues Related j
to Nuclear Power Plants, (NUREG-0410).
4.
Interim Technical Report on %'R Feedwater and Control Rod Drive Return Line No==le Cracking (NUREG-0312).
[
6-5 g
r_m
5.
A Technical Update on Pressure Suppression Type i
Containments in Use in U.S. Light Water Reactor Nuclear Power Plants (NUF2G-0474).
6.
Before Licensing Floating Nuclear Power Plants, Many o
Answers Are Needed, Report to the. Congress by The Comptroller General (EMD-78-36).
7..
Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping (NUREG-0313).
8.
SECY-78-137, March 7, 1978 letter from E. Case to NRC Commissioners.
t t
- 5.
SUMMARY
AND CONCLUSION
.i In reviewing the nuclear industry record and newly available documents over the past two years, it becomes obvious that inclu-si.on of larger accident sequences is essential to more accurately assess risk imposed by the Shoreham plant.
The continuing problems of vessel and pipe ' cracks, inadequate verification of critical system designs, and lack of resolution of ATWS, coupled with erosion of confidence in the Reactor Safety Study make such accident sequences credible events that must be evaluated.
Th Applica$1t and Staff must be required to demonstrate the adequacy of their consideration in the design of this plant.
6-6
.n
l e
CONTENTION 7a NUCLEAR SYSTEM DESIGN DEFICIENCIES 1
1.
STATDENT OF CONTENTION Intervenors contend that the Applicant and Regulatory i
Staff have not adequately demonstrated that the Shoreham nuclear system meets the requirements of 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, with regard to the design adequacy of the following system and response characteristics and/or criteria:
1.
Criterion 3 requires that systems "shall be designed and located to minimize" probability and effect of fires and explosions.
This has not been adequately demonstrated.
(>.
ii.
Criterion 19 requires equipment be outside the control room with the "provided
\\.-
design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in safe con-dition."
This capability has not been adequately demonstrated.
iii.
Criterion 10 requires the reactor core control design to assur.e appropriate margin of the fuel cladding be maintained as one of the multiple barriers.
Lack of fine motion control of the control rod drives, and its potential to cause cladding damage has-not been demonstrated to be in compliance with this requirement.
~
iv.
Correction of the anticipated transient without
(.*
scram (ATWS) and the end of cycle (EOC) scram reactivity problems has not been demonstrated, nor has it been demonstrated that adequate solu-tions may be available and/or implemented to bring the system into compliance with Criterion 20.
v.
Criterion 25 requires reactivity control syste=s be designed to assure that fuel design limits are not exceeded for any single malfunction such as accidental withdrawal of control rods or equip-ment failure.
Adequacy of the design of the system which controls rod sequence has not been demon-
. strated to be in compliance with this requirement.
71
_ _m.m e
vi.
Criterion 61 requires fuel storage systems shall be designed "with a capability to permit appro-priate periodic insp"ection and testing of components important to safety.
Given the current uncertainty regarding reprocessing and long term disposal of irradiated fuel, compliance of Shoreham fuel storage -
facilities with this requirement has not been adequately domonstrated.
vii.
Criterion 62 requires criticality of fuel in the storage and handling system be prevented by physical systems or processes.
Possibility of criticality in the new fuel storage system as a result of reactivity addition by fire protection fogging systems has not been demonstrated to be in compliance with this require-ment.
2.
CONTENTION CHRCNOLOGY This contention, along with all other currently active c
County cententions, was originally described in County's Amended Petition to Intervene, dated September 16, 1977.
3.
BASIS FOR CONTENTION l
The regulatory basis of each of the 'seven parts of this concention is specifically identified in the wording of the contention above.
The following additional information is provided to further expand and explain the reason for concern.,,
i.
Cri'terion 3
~ Fire Protection.
This criterion requires that structures, systems, and components important to safety be designed and located to minimire, consistent with other safety requirements, the probability and effect of fires and explosions.
The primary basis of concern at the Shoreham plant is the fact that this system design predates the Browns Feu-y Fire experience and is therefore subject to question as to adequacy of design in this regard.
7-2
e The fact of failure of two subsequent fire separation tests conducted for che NRC calls into the question the adequacy of the design criteria used.
In addition, the review of station fire detection and fire fighting systems is incomplete at this timd.
)
s 1
11.
Criterion' 19 - Control Room.
Criterion 19 requires among other things that equipment at the appropriate locations outside the control room shall be provided 8
with the design capability for prompt hot shutdown of the reactor and subsequent cold shutdown.
The des-s., /
cription in the FSAR regarding the capability provided ht Shoreham to safely shut down the reactor by use of the remote shutdown panel is cursory and incomplete.
Question exists as to the adequacy ofothe procedures 4
and as to complete consideration of possibility of failures due to systems interactions.
The questien of system interaction in nuclear power plants is addressed in NUREG-0410 wherein Task Action Plan A-17 has been
[,
approved.
Schedule for completion of this task origi-
~
nally was December 30, 1978, but as most of these action plans have been slipping in schedule, it is unlikely that this commitment will be met.
Until assessment of system interaction is provided and complete identifi-cation andl esting of the remote shutdown procedure is t
provided, the adequacy of the Shoreham design to meet this criterion is in doub.
t 7-3
iii.
Criteri'on 10 '~Rea~ctor Design.
This criterion requires that the reactor core, including its control system, shall be designed with appropriate margin to assure that specified fuel design limits are not exceeded e
i i
during any condition of normal operation, f acluding the effects of anticipated operational occurreaces.
The BWR control system has been found to be inadequate with regard to this criterion in that control rod motion i
i during both normal operation and anticipated operational occurrences, results in local power level changes which are detrimental to the fuel cladding, resulting in untimely deterioration of one of the fission product' barriers.
Procedural requirements have been instituted at reactors to overcome this design shortcoming, but the long-term fix of this problem, the application of.
fine motion control rod drive systems, has yet to be implemented.
Other design fixes may be possible to overcome this deficiency, but the fact remains that Reactor Design Criterion 10 is not being met.
)
iv.
Criterion 20
' Protection' System Functions.
This criterion requires that the protection system shall be designed to' initiate automatically the operation of appropriate systems including the reactivity control systems to assure fuel design limits are not exceeded as a result of anticipated operational occurrences.
7-4
This has been interpreted to require that anticipated l
transients without scram (ATWS)~ must be ' considered and appropriate designs provided.
WASE-1270, in 1973, concluded that as more reactors are built and operating, safety improvements will need to be inplemented to reduce the probability / year of an ANS event and/or-mitigate the consequences.
When WASE-1400 was published (1975) and showed ANS l
as one of the primary causes of accidents---particularly- [
in BWR's(1), they assumed 10 transients per year and G
ANS probability of about 10-5/ reactor year for BWR's, less for a PWR.
j The..recently issued NRC report on AWS, NUREG-0460(2),
goes beyond WASH-1270 and' WASH-1400 by refining the estimated values for ANS events to an average of 6 transients per year (8 for a BWR) and 3x10-5/ demand for the probability of failure to scram, leading to an ANS probability of 2x10~4/ reactor-year (3)(those pro-l i
ducing significant consequences).
They set an objective 1
of 10-6/ reactor-year as an upper limit of probability r
of an ANS event.
l There are several factors which will make BWE's likely j
to experience more transients than other plants.
1 (1)
WASH-1400, Appendix I, Figure 4-11, 4-12.
l
\\
i (2)
Anticipated Transients Without Scram For Light-Water Reactors, NUP2G-0460, April 1978.
i (3)
Ibid, Vol 1, page 29.
l I
7-5 i
l l
Statistically, as shown in NUREG-0460, plants of new design experience more transients than older designs (4) l Since SNPS has some new systems' (it is one of the lead Mcrk II containment plants), it is likely to fall into this category.
It seems to also be the case that BWR's experience more transients than PWR's. (5)-
This frequency for BWR's is quoted in NUREG-0460 as 8/ reactor-year vs the average of 6/ reactor-year for all reactors.
The GE design for SNPS uses the old relay-type safety system to initiate a scram signal.
This system is subject to common cause failures including those identi-fled in NUREG-0460{6)
The control rods and drives for SNPS are basically the same ones GE has used for 15 years.
They have redundancy but no ' diversity.
They are clustered and rely on bank actuation of group of rods for scram.
These factors make them subject to
~
common mode failures both inside the reactor (the control rods) and outside (the rod drives).- GE's cal-culation of scram failure probability is listed as
" proprietary" but generally, BWR's are better than PWR's (4)
Anticipated Transients Without Scram For Light-Water Reactors, NUREG-0460, April 1978, Vol.1, page 12 (5) Ibid, Vol 1, page 13.
(6) Ibid, Vol 1, page 17.
t 7-6 A
here.(7)
However, it is not clear if their calcu-lation is for their improved design or the older i
design used on SNPS.
The SNPS design has only one j
system for Standby Liquid Control (SLC) and this l
system (even if improved) is not given much credit for mitigating the effects of an AWS event.(0)
The SNPS design also includes an automatic trip of the recirculation pump on faae turbine trips.
This, by itself, is evaluated in NUREG-0460 as inadequate to prevent a core melt.(9)
The FSAR for SNPS also in-I cludes drawings for a Prompt Relief Trip System (10)
(
which the NRC has found to be an unacceptable solution.
In addition to the above, many of the GE-proposed modifications for AWS do not appear to resolve the problem for several anticipated transients and subsequent failures to scram.(11)
(7)
NUREG-0460, Vol. II, Appendix II, page II-106.
(8)
Ibid, Vol. II, Appendix XVI, page XVI-62, 63.
v (9)
Ibid. Vol. II, Accendix XVI, page XVI-62, 63.
(10) FSAR, Vol. 10, Figures 7.6.1-4, 1-5, & 1-6.
(11) GE-proposed modifications are called " proprietary" (NUREG-0460,
- p. XVI-76) but their failure to miti. gats ATWS is shown in the text of NUREG-0460.
(Ref to pages XVI-62, -63 and -77).
h 7-7 r
In summary, the present and proposed designs !for 1
l SNPS would not survive many of the proposed ATWS 2
sequences, potentially resulting in a core melt and/or subsequent contahent failure with the result of radiation releases similar.to those evaluated in 5
WASH-1400.
Although the ACRS, NRC, reactor vendors, and consultants continue to meet on the ATWS problem, and it has been given a top category ranking in the generic issues list (12) it is not clear when or if the solution to ATWS will be implemented on older designs such as BWR-4's and 5 's.
v.
Criterion 25'
' Protection System Requirements For Reactivity Control Malfunetions.
This criterion requires that protection systems shall be designed to assure that fuel design limits are not exceeded for any single malfuntion of the reactivity control systems such as accidental withdrawal of control rods.
This function is to be prov*.ded at the Shoreham plar.]
by the use of the rod sequence control system (RSCS).
The description of the RSCS is found in'the FSAR on page 15.1-32(b).
The description is simplistic and cursory and states only that the RSCS " prevents the selection of an out 'of sequence control rod."
The RSCS was an added on element in the Shoreham control system design and has never been fully qualified by l
(12)
NUREG-0410, Item A-9.
7-8
i j
testing with all interconnecting systems, and under 2
the conditions of complete interaction of the con-i!
i~
trol system.
Consequently, its ability to perform its intended function is in doubt and additional proof of its adequacy should be presented by Applicant i
and/or Staff.
vi.
Criterion 61 - Fuel Storage.
This criterion requires that the fuel scorage and handling systems shall be designed to assure adequate safety under normal and 4
(
postulated accident conditions, and that the system shall be designed with the capability to permit appro-
{
priate periodic inspection and testing of components important to safety.
The Shoreham fuel storage facilities i
are believed to be inadequate from the standpoint of ability to perform appropriate periodic inspection and testing of spent fuel storage' racks and attachments.
These components were originally designed for short-term I'
'~
~
service, as it was intended that spent fuel would be s
i
.ll shipped from the site for fuel reprocessing after a short period of cool-down fuel storage.
With the o
current situation requiring that spent fuel be stored at the operating plant storage pools almost indefinitely, fuel storage racks are to be expected to be in service
'1 on a long-term basis and appropriate periodic inspection i.
i -
schedules should be identified.
This has not been done.
7-9
.-,-.,.-e,,,e
--.- c
vii.
Criterion 63 - Prevention of Criticality in Fuel Storage and Handling..This criterion requires that criticality in the fuel storage and handling systems shall be prevented by physical systems or processes.
Concerni. exists as to the adequacy of the Applicant's criticality evaluation concerning the possible acci-dental criticality of new fuel in the new fuel storage racks as a result of the partial flooding or high-density fog spray of fire protection systems.
The accident analysis of this event contained in Section
\\
r 15 of the FSAR does not include the assumptions used ',,)
in calculating sn L effect and since it has been a subject of concern and uncertainty for a number of years, particularly during the design phase of the
. plant, additional evaluation of this particular problem is required.
4.
REFERENCES Section i i
a.
Regulatory Guide ~1.120, " Fire Protection Guidelines for Nuclear Power Plants," Rev.1, November 1977.
b.
Regulatory Guide 1.75, " Physical Independence of Electrical Systems," Rev.
2', September 1978.
c.
Branch Technical Position APCSB 9.5-1, " Guide for Fire Protection for Nuclear Power Plants Docketed Prior to July 1,1976'.', including Appendix A, August 1976.
l l
d.
NUREG-0050, " Recon:mendations Related to Browns Ferry Fire,"
l report by Special Review Group, February 1976.
e.
Letter, Feit (NRC) to Tong (NRC), August 5,1977, includes Trip Report of Sandia Fire Test Review.
7-10
~
f.
Letter, Snyder (NRC) to Tedesco (NRC), October 6,1978, re:
Underwriters Laboratory Fire Tests on September 15, 1978.
g.
Letter, Denton (NRC) to Gilinsky (NRC), November 2,1978, subject:
NRC Fire Protection Research Test.
Section 11 a.
NUREG-0410, "NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants," January 1978.
b.
Branch Technical Position APCSB 9.5-1 (see (i) above).
Section 111 a.
FSAR - SNPS, Section 7, Volumes 8-10.
A Section iv a.
WASH-1400 (NUREG-75/014), " Reactor Safety Feudy", U.S. NRC, October 1975.
b.
NUREG-0460, " Anticipated Transients Without Scram for Light Water Reactors," April 1978 c.
FSAR - SNPS, Volu=e 10.
d.
NUREG-0410, January 1978 (see (ii) above).
Section v a.
FSAR - SNPS, Section 7, Volumes 8-10.
i Section vi a.
FSAR - SNPS, Volumes 5 and 6.
i Section vii i
a.
FSAR - SNPS, Section 15. Volume 13.
7-11 i
\\
l
= _ __. _
_=
5.
SUMMARY
AND CONCLUSIONS In summary, the identified critical centrol systems have not been demonstrated or described to assure compliance of the Shoreham plant with appropriate design, criteria.
This is a result of rapidly changing design philosophy during the lengthy period of this reactor project.
The reactor vendor has advanced the product line to a newer version than Shoreham represents,
and their generic design verification efforts may not be applicable to the Shoreham design.
These contentions represent serious potential deficiencies to safe plant operation, and should be thoroughly reviewed and considered during the course of the. operating license hearing
~
process.
4 O
l 7-12 a
CONTENTION Ba HUMAN ERROR i
t 1.
STATEMENT OF CONTENTION 8a.
Intervenors contend that the Applicant and Regulatory Staff have not demonstrated that the impact of human error has been analyzed as required by 10 CFR Part 50, Appendix A, Criterions 19, 20, 22, and 29 with regard to:
Equipment arrangement in the control room i
and on the control consoles.
ii.
Automation and computer control of emergency equipment.
3 2.
CONTENTION CHRONOLOGY This contention, along with all other currently active County contentions, was originally described in County's Amended Petition to Intervene, dated September 16, 1977.
3.
BASIS'FOR CONN.nuGN The control room arrangement a't SNPS is of the large open-L configuration which results in long panels and difficult visi-c(-)
bility from one corner of the control panels to the opposite corner.
The many thousands of components (with large numbers of them being identical or similar devices) make distinction of specific functions at a distance nearly i=possible.
The long distance between opposite ends of the "L" (FSAR Figure 3.8.4-6) could mean a delay in reactor operator response to developing situations.
When operator action is taken, the 1
multitude of similar devices and components lends itself to improper actions as a result of human error.
(See EpRI N?-
l l
30 9-S'd.
Such a design is contrary to the requirements of l
o,
Criterion 19 and 22.
The. design also allows for separate Protection System Divisions to be brought within a very small distance of one another in bhared control devices.
(FSAR, Section 7.3.3.1.2.17g(i), page 7.3-68).
This is not in conformance with criterion 22 since a single fire can impact both divisions at the device.
The system is declared to be " tolerant to.any single raceway fire" (FSAR, page 7.3-7.I) which in the main control panel may involve a total section of the panel, thus impacting shared devices.
Criterion 22 also requires inter alia that "the effects of..... maintenance testing.......do not result in loss of the ' j protection function."
The SNPS design permit several manually initiated actions to render emergency systems inoperative'(FSAR, pages 7.3-77' and 7.3-78).
These and other actions could occur during test and/or mMntenance operations and are subject to human error in operation and detection.
These actions are generally under only supervisory control and not all are annunciated in the control room.
In fact, the FSAR takes ex-ception to the requirement in IEEE 279 for continual monitori:]
and control room indication of inoperability" of protection functions (FSAR, page 7.3-78).
Diversity, in the sense used in Criterion 22 is essentially absent in many portions of the Protection Systen.; thereby allowing human error to impact the reliability of the safety function.
The use of similar or identically designed components for initiation, monitoring, displaying, or actuation of safety 8-2
' " ~ - - - * - - * -
O functions is clearly lacking in diversity.
Examples are the actuating relays in the protection system, the control switches on the control panels, and the control rod drive devices.
Criterion 19 also requires remote shutdown capabilities in the event the control room is uninhabitable.
- However, there is insufficient information to show the control cables to and from the remote panels do not pass through the cable spreading room.
Layout and design are not in conformance with the latest Regulatory Guides for separation.of redundant safety divisions (see $5AR, Figure 3.8.4-6 and Reg Guide 1.120,
\\- -/
Revision 1), and as such are vulnerable to human errors which could lead to co= mon-cause failures such as fires.
4.
RE?ERENCES 1.
FSAR - SNPS, Volumes 6, 8, 9, and 10.
2.
CFR, Title 10, Part 50, Appendix A.
3.
Regulatory Guide 1.120, Revision 1, Nov. 1977.
4.
EPRI NP-309-SY, Nov.1976, Human Factors Review of Nuclear
' N Plant Control Room Design.
\\.s 5.
SUMMARY
Si CONCLUSION The SNPS control room and cable spreading room designs are old and cumbersome designs with weaknesses which make them vulnerable to human error.
They are also not in conformance with the latest Regulatory Guides in some areas.
These weak-nesses could lead to loss of control, accidents, or complication of anticipated transients / accidents.
Ecwever, to pursue these 8-3
_. _. 7 ;, _
issues will be coscly and time-consuming and will be countered by arguments that the new criteria were not in effect for this h'
plant design.
Therefore, it is recommend that this issue be 1
carried out at lower priority.
I 2
i
[
b t.
l l
l 4
1 1
, _.)
I l
i i
a e
i i
8-4 l
CONTENTION 9a & b REACTOR CONTAINMENT SYSTEM LOCA RESPONSE 4
1.
STATEMENT OF CONTENTION 9a.
Intervenors contend that the Applicant has not adequately demonstrated that the Shoreham reactor containment system meets the requirements of 10 CFR Part 50, Appendix A, Criteria 4,16,.
50, 51, and 52 with regard to:
1.
Response to forces generated during suppression pool LOCA dynamics, such as, but not limited to, pool swell, vent clearing, lateral vent. loads,
. chugging, seismic slosh, etc.
ii.
Ability of the drywell/wetwell membrane to with-
~
stand the LOCA induced transient load it may s.
experience.
iii.
Ability of the containment pressure boundary to prevent fractures that may be induced by pressure or impact loads from transients, dynamics, or missiles.
iv.
Capability to test the leakage rate of the drywell/
wetwell membrane and other critical parts.
9b.
Intervenors further contend that.the Applicant has not developed adequate experimental data to verify the containment design in accordance with requirements of 10 CFR Part 50, Appendix B, Criteria III and Criteria XI, with specific regard to:
1.
Testing "under the most adverse design conditions."
ii.
Performance of tests under suitable environmental.
conditions.
iii.
Documentation and evaluation nf test results, iv.
Use of tiest data developed under a non-controlled test program (foreign).
2.
CONTENTION CHRONOLOGY This contention, along with all other currently active County contentions, was originally described in County's l
Amended Petition to Intervene, dated September 16, 1977.
9-1
3.
BASIS FOR CONTENTION 1
This contention, along with C,ontention 10a and b, focuses on the issues of the design adequacy of the Shoraham Mark II Containment system to withstand the postulated effects of a LOCA event separately, and in conjunction with the transient l
effects of SRV blowdown.
Contention 10a and b addresses the '
4 ability of the system to withstand ~ roads imposed by SRV blow-I down during operational transients and during postulated LOCA.
Since these two conditions are inextricably intertwined, their i
bases will be addressed in' combination in this discussion.
For
/.
reference to specific Contention 10a and b wording, see that l
section of this report.
1 The regulatory basis of both Contentions 9a and 10a is found
{!
in 10 CFR 50, Appendix A, Criteria 4, 16, 50, 51, and 52.
Cri-terion 4 specifies that structures and systems important to 4
i safety shall be designed to accomodate the effects of and to be compatible with the environmental conditions associated with both normal operation and postulated accidents including loss of coolant accidents.
Q-Criterion 16 specifies that the containment design shall establish an essentially leak-tight barrier against the uncon-trolled release of radioactivity.-
Criteria 50, 51,. and 52 address specific details of contain-ment design to comply with LOCA response, fracture prevention of the containment, and capability for leak testing.
Substantial concern about the design adequacy of the Mark II Containments to meet these requirements is a result of inade-quate design verification / testing during the conceptual design 9-2
j.
phases.
The result of this incomplete design verification 1
program was the establishment of' design parameters with inade-quate or complete lack of design loads being specified for j
two very important plant loading conditions.
The two conditions
.i include definition of loads occurring to the containment during the loss of coolant accident and ' loads occurring at initiation of safety relief valve discharge from the reactor to the suppression pool.
NRC's recent report, NUREG-0474, A Technical Uodate on a
.f.,
Pressure /Sucoression Tvoe Containments in Use in U.S. Light-Water Reactor Nuclear Power Plants, July 1978, was issued at the request of Congress to provide a summary update of reactor containment design problems.
The lack of attention to these critical loads is well described in this report as follows, quoting from pages v and vi:
"An evaluation of the Marid III pool swell data indicated the need for a re-assessment of both Mark I and Mark II containment designs.
These facilities had not explicitly considered the pool hydrodynamic loads associated with a postu-lated loss of coolant accident in their designs."
- s.,
"Until the early 1970's, the only significant hydrodynamic loads considered in the containment design were those with a postulated LOCA.
During this time period, safety relief valve induced loads were considered to be small and secondary."
The situation is further described in General Electric's i
report NEDO-21061, Mark II Containment Dynamic Forcinz Functions Information Recore, September 1976, where on page 5-2 it is stated 9-3
4 l
j "In the past, LOCf loads have been assumed to consist of a gradual pressure rise in the j
suppression chamber and/or dry well, in j
addition to a much slower temperature rise.
......both effects were regarded as quasi-static."
What the foregoing statements mean is that the original design basis of the Mark II Containment did not include consideration i
of the effects of the LOCA hydrodynamic loads nor of the safety relief valve discharge dynamic loads.
It was.not until 1974 when.
the NRC began asking questions concerning the Mark III testing
)
~ results that General Electric and the Mark II utilities bdgan to face up to the problem.
It should be emphasized that the loads in question are substantial and significant.
Dynamic load definition is a com-plex design process and it is difficult to summarize the magnitude and/or significance of the Mark II design omissions.
Ecwever, it 4
may be easiest to indicate the magnitude of the omissions by com-paring some of the LOCA or SRV loads now being considered to the original " quasi-static" loads for which the containments were
~
originally designed.
NEDO-21061, the Mark II Dynamic Forcing Function Information Report, contahn a Table 2-1, page 2-2, which summarizes the LOCA-related and SRV-related loads magnitudes that have been predicted by analytical model and which are currently exp2cted to be verified in a number of ' test facilities all over the world.,The first load listed in the table is the LOCA-related contain=ent clearing transient load, which was specified as 33 psi.
This is a containment wall pressure load which is al=ost equal in 9-4
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. _i__
_1 i magnitude er the " quasi-static" design containment pressure load of 36 psi.
From the same table it can be seen that the containment wall loading imposed by the discharge of a safety relief valve in asymmetric condition is 23 psid and the load could run as high as 34.6 psid for all valves discharging with Ramshead mitigation devices.
These dynamic loads.in some cases are directly added to the calculated " quasi-static" LOCA load and thus it can be seen that the dynamic loading conditions may welb. increase the contain-ment load by a factor of two or more.
Table 2-1 also identifies safety relief valve-related quencher and Ramshead tie-down loads.
These range from 8,000 pounds horizontal to 166,000 pounds vertical for the quencher and 100,000 pounds horizontal to 70,000 pounds vertical for the Ramshead.
For comparison purposes, these can be compared to design earthquake loads listed in Table 3-16, page 3-111, where it is found that 1.25g earthquake loads range in general from 1,500 to 4,925 pounds.
It is obvious that the dynamic loads now deemed necessary for inclusion in the Mark II-Contaic=ent design are substantial, significant, and unfortunately, as yet unconfirmed.
Despite the confidence stated in NUREG-0474 that the Mark IUcontainment loads will be ' adequately considered, substantial work remains to be done to verify the adequacy of the current load assumptions.
Some of the key design verification and design review steps remaining to be completed are:
1.
Completion of the Caorso tests in Italy to identify Mark II safety relief valve load functions, scheduled l
9-5
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for mid to late 1978 but complete review of results i
will take substantially longer.
Completionofthefull-sAaleMarkIItestsscheduled 2.
in Japan to evaluate Mark II multi-vent loads.
Con-struction of the test facility was begun in summer of 1978 and testing is scheduled to begin in April 1979.
3.
Issuance of the NRC's Safety Evaluation Report, a key tailestone in the design review insthe licensing process is not scheduled for the Mark II program until sometime in 1980.
Because of the ongoing design verification program, it is quite likely that substantive containment changes will be necessary and/or desirable.
An example of the " dynamic" nature of the design evaluation of dynamic loads is found in the NRC's November 17, 1977 letter filed in Docket No. STN-50-447, which contains a summary of a meeting held with General Electric to discuss alternatives for resolving safety relief valve design deficiencies.
This deficiency has to do with the underestimation of the n-her of safety relief valves which might open and which underestimated the loads this effect would apply to the containment.
The problem applies not only to Mark II Containments, but to Mark I and Mark III.as well.
The essence of this underestimation can be found on page 3 where the following is pointed out:
1.,
The nearly 1007. increase in the loads would neces-sitate additional snubbers (to the piping).
'9-6
2.
GE stated,that the reactor internals would probably l
have to be replaced since it was not a simple matter to stiffen them.
The estimated replacement cost would be one to two million dollars per plant plus one or two years delay.
3.
The investigation had been limited to the free-standing st' eel Mark III Containment.
However, the effect of this increased loading condition on the Mark III coi -
tainment would be to require that the thickness would have to be increased to 2 3/4 inches and the estimated increase in cost would be ten million dollars per plant with delays ranging from two to five years.
-(
The Shoreham Containment dc. sign verification is subject t:o_all these uncertainties.
There are possibly more that are unique to Shoreham.
The latest status report of discussions between the Mark II Owners Group, and the NRC Staff is found-in the Nov.1,1978 memo written by Staff's I.A. Peltier.
This memo documents the Oct.: 19, 1978 Staff-Mark II owners Group meeting to discuss Staff's Acceptance Criteria.
The Owners Group took exception to 21 of 2fiuof 5the Staff's load definition criteria.
Additionally, copies of transpatencies used in the meeting indicate that the Shoreham Pool Swell condition is " plant unique" and that the Acceptance Criteria impact on Shorehr -
could result in a one-year delay and a cost of $20-150 million.
The proposed Acceptance Criteria in most cases requires the use of changes in analytical techniques to make up for the inadequate 9-7
design loads originally specified.
The proposed square root of the sum of the squares method for evaluating dynamic loads has been identified in licensing literature 'as a " pivotal" piece of design methodology.
What is being proposed is a method of analyti-cally reducing the peak loads by as much as 3'0-507. rather than i
design the plant as it would be under normal conditions.
This is in conflict with the approach proposed by the NRC in evaluating the new Mark III Contahment in NUREG-0484, where only the reactor coolant pressure boundary and not the containment itself will be permitted to use SRSS.
Discussions with NRC personnel disclosed that additional meetings were scheduled for ')
Nov. 14, 15, and 17, 1978.
Results of those meetings are not yet available to MEB.
In addition, a three-day meeting to discuss these issues with the ACRS is scheduled for Nov. 28-30, 1978 in San Francisco.
A copy of the tentative agenda (attached) indicates extensive review of these factors is continuing.
The Shoreham containment appears to be headed towards evaluation in accord with non-standard, non-conservative techniques which are being applied differently so as to circumvent the regu- ]
1ations.
A thorough review is not yet possible, however, due to the incomplete nature of the ongoing Mark II program.
Contentions 9b and 10b address the issues of the adequacy of the Mark II Containment LOCA and SRV effects design verification test program.
The basis of contention was well s" arized in the County's Jan. 31, 1978 Response to Applicant's Second Interrogatories, repeated here for ce=pleteness.
9-6
Applicant, in conjunction with the Mark II Owners Group has chosen to use " experiments" performed in the tests conducted by various suppliers and organizations in Sweden and Germany (and now in Japan) to verif7 the design basis for the Shoreham plant.
It has not been demonstrated that such testing has been performed "under the most adverse design conditions" nor that the performance of the tests was (will be) done "under suitable environmental con-ditions," or that documentation and evaluation of test results has not been demonstrated to have been conducted in accordance with U.S. quality assurance criteria as required by 10 CFR 50, Appendir B, Criteria III and II.
4.
RFERENCES 1.
NUREG-0474, A Technical Update on Pressure Suppression Type Containments in Use in U.S. Light Water Reactor Nuclear Power Plants, USNRC, July 1978.
2.
NUREG-0484, Methodology for Combining Dynamic Responses.
~
3.
NEDO-21061, Mark II Containment Dynamic Forcing Functions Information Report.
4.
Mark II Containment Supporting Program Document St" mary, 12/77 (Transmitted with R.K. Hoefling's 12/21/77 letter to
~,
D.G. Bridenbaugh).
5.
Nov.1,1978 memo by I.A. Peltier, subject:
Meeting With Mark II Owners Group to Discuss Staff's Mark II Containment Acceptance Criteria, Oct. 19, 1978.
~
6.
Nov. 28-30, 1978, Proposed Presentation Schedule - ACRS Meeting of the Fluid Hydraulic Dynamic Effects Subcommittee, San Francisco, California.
7.
NUREG-0410, NRC Program For Resolution of Generic Issues Related to Nuclear Power Plants.
)
l 9-9
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5.
SUMMARY
AND CONCLUSION It is clear that the adequacy of the Shoreham containment system to withstand both expected transient and LOCA loads has not been demonstrated by the Mark II generic effort.
Addition-a31, Shoreham unique problems appear to exist and to be less 7
than candidly described in the open documentation.
../
O e
f 9-10
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A
[-
f NOV 2 0 378 Proposed Presentation Schedule
~
i ACRS Meeting of the Fluid Hydraulic Dynamic Effects Subcomittee San Francisco, California November 28, 1978 9:00 AM 9:00 - 9:15 I.
ACRSOpeningComments-(M.Plesset) 15 min.
9:15 - 10:45 II. Methods for Combining Loads (NRC/J. Knight) 90 min.
10:45 - 11:45 III. Mark II 0.G. Comments on load Combinations Acceptance Cr,iteria (Mark II 0.G.)
,60 min.
11:45 - 12:45 LUNCH 12:45 - 3:15 III.
(Continued) Mark II 0.G.' Coments on Loard Combinations Acceptance Criteria (Mark II 0.G.)
150 min.
November 29, 1978 9:00 AM 9:00 - 9:20 I.
Introduction (NRC/R. Tedesco) 20 min.
9:20 - 10:50 II. SRV Related Hydrodynamic Loads A.
SRV Loads Overview (NRC/T. Su) 30 min.
B.
Air Clearing Loads'
~
1.
Acceptance Criteria (NRC/T. Su) 20 min.
a 2.
Confirmatory Programs (NRC-MIT/P. Huber) 30 min.
C., Pool Temperature Limits (NRC/T. Su) 10 min.
^
~10 11:00 BREAK 11:00 - 11:45 III. LOCA Related Hydrodynamic Loads A.
LOCALoadsbyerview(NRC/C. Anderson) 45 min.
11:45 - 1:00 LUNCH 1:00 - 2:45 III. (Contirhi'ed) CCCA Rel'ated Hydrodynamic load B.
Pool Swell (NRC-BNL/C. Economus) 30 min.
9-11
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1.
Pool Swell Elevation and Velocity i
, 2.
Asynnetric Loads C.
Impact Loads (NRC/BNL/G. Maise) 45 min.
1.
Small Structures 2.
Gratings D.
Steam Condensation Loads 1.
VentLateralLoads(NRC-BNL/C.Econoras) 15 min.
2.
Chugging - FSI Considerations (NRC-Princeton /R. Scanlon) 15 min.
1Y. LOCA/SRV Submerged Structure Drag Loads (NRC-Princeten/G. Bienkowski) 75 min.
i
]l A.. Jet Loads B.
Air Bubble Drag Loads l
4:00 - 4:10 BREAK 4:15-4:35 Y.
RelatedNRCResearchPrograms(NRC/R.Cudlin) 25 min.
4:35 - 4:45 VI. Conclusions (NRC/W. Butler) 10 min.
l
'4:45 - 5:00 YII. Coments (Mark II OG) 15 min. l November 30, 1978 9:00 AM 9:00 - 10:00 Mark II 0.G. Con:nents on 'Hydrodyna:nic Loads a
o Acceptance Criteria (Mk II OG) 6C ).
10:00 - 10:10 BREAK 10:10 - 11:30 Mark II SRV Test Programs (Mk II OG) 80 min.
A.
W TECH B.
KWU C.
Zinr.er Inplant Tests i
l f
9-12
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a 3
11:30 - 12:30 LUNCH I
12:30 - 3:00 Intermediate Program Tasks (Mk II OG) 150 min.
A.
Dynamic Lateral Loads B.
Chugging Leads 1.
Ringout Removal 2.
Multivent Model 3.
Test Programs C.
Submerged Structure Drag Tasks 1.
Analytical Programs 2.
Test Programs D.
SRV Tasks 1.
New Four-Arm Quencher Lead Methodology 2.
Temperature Limits 3:00 - 3:15 ACRS Closing Comments 15 min.
~
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t CONTENTION 10a & b STRUCTURAL RESPONSE TO TRANSIENT CONDITIONS 1.
STATEMENT OF CONTENTION 10a. Intervenors contend that the Applicant has not demon-strated that the Shoreham reactor containment and pressure vessel supporting structure meet the requirements of 10 CFR Part 50, Appendix A, Criteria 4,16, 50, and 51 with regard to:
1.
Structural loadings and response to the "condi-tions associated with normal operation" such as transient and extended blowdown of safety-relief valves into the suppression pool.
ii.
Maintenance of "an essentially leak-tight barrier" during such expected conditions.
iii.
Adequacy of the design to ensure,'with sufficient margin, that the containment strucute can accommo-date the simultaneously applied loads of transient and LOCA events without exceeding. established leakage limits, iv.
Adequacy of the reactor foundation to withstand, without amplification or failure, the transnission of transient-generated loads in the suppression pool.
10b. Intervenors further contend that Applicant has not yet developed adequate experimental data through testing in accordance with requirement of 10 CFR Part 50, Appendix B, Criteria III and XI to assure the verification of the design through a properly controlled test program.
2.
CONTENTION CHRONOLOGY This contention, along with all other currently active County contentions, was originally described in County's Amended Petition to Intervene, dated September 16, 1977.
10-1
4 t
3.
BASIS OF CONTENTION See Contention 9a and b.
. i 4.
REFERENCES See Contention 9a and b.
5.
SUMMARY
ANL CONCLUSIONS See Contention 9a and b.
e e
O 6
10-2
.o.
CONTENTION lla EMERGENCY CORE COOLING SYSTEM i
1.
STATEMENT OF CONTENTION l
lla.
Intervenors content that the Applicant and Regulatory Staff have not adequately demonstrated that the Emergency Core Cooling System (ECCS) for Shoreham meets the requirements of 10 CFR Part 50.46 and 10 CFR Part 50, Appendix K with regard to:
1.
Peak cladding temperature.
11.
Counter Current flow disruption.
1 111.
Flow blockage and effect of bypass flow, iv.
Effect of metal water reaction.
v.
Cladding performance.
2.
CONTENTION CHRONOLOGY This contention, along wich'cll other currently active County contentions, was originally described in County's Amended Petition to Intervene, dated September 16, 1977.
Companion Contention 11b was dropped by County as indi-cated in County's November 10, 1977 Amended Contentions and Legal Arguments document.
3.
BASIS FOR CONTENTION Despite efforts to the contrary, there is still a lack of complete proof in full scale facilities and LOCA en-vironments, of the capability of ECC systems and the ad-equacy of the predictive models.
Although questions remain 11-1
,, c;,,
in each of the areas of this contention, the major un-certainty lies in the core spray capability to adequately cool the core and the accuracy of the computer codes to predict the performance of core spray and other ECCS.
The possible core spray inadequacies are described in detail in Task Action Plan A-16(1), Steam Effects on BWR Core Spray Distribution.
In this Task Action Plan, it states that "There is no experimental verification that the ' air mock-up of steam environment' tests that were conducted for BWR 2 l,
through BWR 5 plants actually predict the spray distribution that would exist in the real steam environment following a LOCA."
A recent meeting with GE and the NRC (1/19/78) aows continuing uncertainty of idependence or separability of thermal and hydraulic effects in core cooling using the core spray system.
This uncertainty will continue until adequate testing is completed and the results are thoroughly evaluated.
The FSAR for SNPS does not provide an adequate description of the full scale multi-nozrle, steam environment, LOCA-q) condition testing of core spray distribution needed to meet the requirements of 10 CFR 50, Appendix K and to resolve the (1)
NUREG 0410, "NRC PROGRAM FOR THE' RESOLUTION 07 GENERIC
. ISSUES RELATED TO NUCLEAR POWER PLANTS", January 19 78
- 4 11-2
e uncertainty of the core spray cooling ability.
Instead, it relies on analytical models to predict the ECCS results for i
s i
different sizes and types of full scale reactors based on tests of small scale models, components and' mock-ups of core sections.
Experts have expressed disagreements on the adequacy of these codes and models at technical meetings such as the July 13, 1977 ACRS subcommittee meeting on ECCS.
The NRC has recently imposed reactor deratings and/or operating restrictions on BWR (and PWR) reactors following discovery of errors in the computer coda.s used for predictier.
of ECCS performance.
In addition they are planning an audit of manufacturers practices and QA related to code development and implementation.
The last such formal audit was in 1974.
The review schedule contained in Task Action Plan A-16 in-dicates completion by December 31, 1979.
It seems apparent that these inadequacies and uncertainties in ECCS apply directly to Shoreham and should therefore be considered in the operating license review.
The interveners concern for th issue is further substantiated by the caveat in Section 8 of 4
the TAP-16 wherein it is stated that if the review or testing causes questions regarding the safety of continued plant operation, they propose to grant exemptions to certain re-quirements in 10 CFR 50.46 so the affected plants can continue l
to operate.
Such an action on SNPS would be in direct opposi-tion to the substance of this contention.
11-3
4 i
4.
REFERENCES 1.
FSAR-SNPS, Section 15 2.
NUREG 0410, "NRC PROGRAM FOR THE RESOLUTION OF GENERIC ISSUES RELATED TO UNCLEAR Pt4TER PLANTS",
January 1978.
3 i
3.
NRC Press Release 78-140, June 14, 1978.
4.
Transcript, ACES Meeting of Subconunittee on ECCS, July 13, 1977.
5.
Minutes of GE/NRC meeting on January 19, 1978, Attached to letter, R. Woods to R. Baer (NRC) dated February 3, 1978.
5.
SUMMARY
AND CONCLUSIbNS There is sufficient uncertainty about the ECCS capability for a BWR, particularly the core spray capability and the computer models, to warrant review of this contention at the operating license hearings for Shoreham Nuclear Power Station.
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i I
i l
9
l
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l CONTENTION 12a l,
NUCLEAR SYSTEM COMPONENT FAILURES / DEFICIENCIES j
1.
STATEMENT OF CONTENTION i
12a.
Intervenors contend that the Applicant and Regulatory Staff have not adequately demonstrated that the Shoreham nuclear system meets the requirements of 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants with regard to the design adequacy of the following system components:
1.
Control rod life.
ii.
Susceptibility of local power range monitors (LPRM) to flow-induced vibration failures.
iii.
Instru=entation set point drift.
iv.
Safety-relief valve reliability and functionability, v.
Ability of containment penetration seals to withstand the design accident conditions.
vi.
Isolation valve configuration and functionability.
vii.
Ability of system valves, pumps, and heae ex-changers to perform throughout the life of the plant.
viii.
Ability of condenser tubes and ' condensate deminera-lizers to protect against injection of sea water into the primary system.
l 2.
CONTENTION CHRONOLOGY This contention, along with all other currently active l
County contentions, was originally described in County's t
l Amended Petition to Intervene, dated September 16, 1977.
The Board ruling in the October 11, 1977 Prehearing Conference was that this contention was acceptable for litigation.
This decisien was confirmed in the January 27, 1978 Board Order.
12-1
,3;,,
1 j
1
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l
{
3.
BASIS FOR C0hrt.MrlON The primary basis of this contention is the poor perfor-mance in general. of the c'omponents identified in light water reactors, and specific problems with these components in BWR's in particular.
County's Response To Applicant's Second Set of Interrogatories identifiad the regulatory basis for the contentien for components listed in the' contention.
This response is re-peated here for purposes of completecess:
COMPONENT 10 CFR Part 50 FAILURE /DEFICIENCT
~Appen' dix A, Criterion' (La) i.
Control rod life.
26 11.
LPRM flow-induced vibration.
4, 12 111.
Instrumentation set-point drift 4, 12, 20, & 21 iv.
Safety relief valve 4, 30 v.
Penetration seals 3, 4, 5, & 50 vi.
Isolation valves 4, 30, & 50 vii.
Miscellaneous valve pumps,.
need further and heat exchangers definition J
viii. Condenser tubes and condensate need further damineralizers definition The following discussion provides additional information on details of the basis of' contention for each of these items in turn and provides a sutstary of status of the on-going review conducted by ME3 for the County.
i.
Control Rod Life.
Criterion 26 requires that suitable and redundant reactivity control system capability shall be provided.
l 12-2
1 The primary basis of carrying this item as an item of con-tantion in the Shoreham hearing is MEB's personal knowledge that BWR control rod life had not been adequately determined i
as of the time that the.Shoreham plant design was finalized.
j Control rod life may be' limited either by nuclear or mechanical design properties.
Baron depletion rates have not been accurately measured and the ability of the control rod stainless steel tubes to perform in the reactor core environment has not been assured.
Information of current control rod life testing should be sub-mitted by Applicant and/or Staff if new information is available.
11.
LPRM Flow-Induced Vibration.
Failure of LPRM's and assc, ciated channel wear was a common problem in the BWR-2,.3, and Modifications have been performed on some reactors and a
4 models.
General Electric is reported to be performing extensive flow-induced vibration tests in their San Jose facilities.
Appli--
cability of design modifications and flow-induced vibration testing.to the Shoreham reactor is uncertain since the Shoreham design contains many of the features of the older plant models
.y and most new testing is specifically addressed to the BWR-6 g l
model.
Additional information is required to be submitted by.
Applicant and/or Staff to determine if this item has been resolved at the Shoreham facility.
iii.
Instrumentation Setpoint. Drift.
10 CFR Part 50, Appendix l'
A, Criteria 4,12, 20, and 21 require erivironmentally suitable i
designs, capable of controlling and/or assuring that core power oscillations do not cause fuel damage,.with appropriate protection system capability and reliability.
Operating reacters of 12-3
I the design vintage of the Shoreham plant have experienced substantial and significant trouble with the maintenance of necessary accuracy required to comply with these regulations.
Accordingly, this system design should be reviewed with that specific deficiency in mind and assurance. demonstrated that this problem has been corrected in the Shoreham plant.
iv. ~ Safety Reli~ef Valves.
10 CFR Part 50, Appendix A, Criteria 4 and 30 require 'that suitable reactor coolant pressure boundary enmponents are utilized with appropriate environmental qualifi-cations.
Substantial problems have been experienced with BWR safety relief valves in the area of reliability and functionabilit:
Specifically, valves have not been found to maintain operability at proper setpoints and valves have frequently stuck open, blowing down the reactor coolant system.
Primary problems have been with Target-Rock SRV's..
A generic item, B-55, Improved Reliability of Target-Rock Safety Relief Valves, has been identified by the NRC.
The Shoreham FSAR appears to indicate that Target-Rock
' ~
valves are not to be utilired at this plant, but no information appears to have been developed describing the reliability and functionability of the valves that are to be used.
NRC Generic Task C-ll, Assessment of Failure and Reliability of Pumps and Valves',would seem to indicate that the NRC considers this to be a general problem area.
Unfortunately, no approved action plan has been developed for the resolution of C-11 and therefore this item has yet to be resolved.
12-4
v.
Penetration Seals.
Ability of the containment penetration seals to perform under long term operating conditions and specifically under the post-accident environmental conditions, e
has long been a concern, particularly with'the vintage of i
penetration seals used in the Shoreham facility.
seals are required to be in compliance with 10 CFR Part 50, AppendixA, Criteria 3,4[5,and50whichhavetodowith design features to be resistent to fire, to perform in the environment expected,' to comply with suitable separation require-i ments, and to assure leak tightness and functionability of the containment system itself.
This concern was confirmed by the /
fact that the NRC has identified penetration seal design and performance 'as a generic problem area and has assigned' a specific task to resolve it.
This task is B-9, Electrical Cable Penetrations 'of Containment, and is identified in the -
unissued generic task problem description document, NUREG-0471.
(This unissued document has receatly been submitted into the public record as part of NRC Staff testimony on Black Fox Construction Permit Proceedings.)
Qualification of the Shoreba.
penetration seals to perform as required should be addressed b Applicant and Staff submittals to assure that ShoIehamis l
not jeopardized by seals of inadequate design or quality.
vi.
Iso 1~ation Valves.
Operating BWR's have experienced a substantial difficulty in maintaining isolation valves in their required leak-tight and operable condition.
This is as required by-10 CFR Part 50, Appendix A, Criterion 4 for environmental adequacy, Criterion 30 for quality of the reactor coolant 12-5
,, 7,,,
_~
1 pressure boundary and Criterion 50 for maintenance of the e
^
containment system leak-tightness and functionability.
Newer t
reactor designs have proposed the use of improved isolation s
' valves or more ' effective leakage control systems, but the L
Shoreham system being generated during an older model era, is
(
of questionable ~ adequacy.
This is verified by the NRC's Generic Task C-ll, Assessment of Failure and Reliability of l
Pumps and Valves, mentioned above, and by numerous failures l
reported in NRC License Event Report.
Assurance tha,e Shoreham l
l will utilize a up to date and adequate design has not been I
demonstrated.
i vii.
Valves', Pumo's', and Heat Exchangers Performance.
This i
general category of component failures and deficiencies was i
originally inserted in the County's contention as a result of l
the knowledge that BWR valves, pumps, and heat exchangers j
had exhibited general poor performance.
As addressed in Section vi above, this fact is attested to by humerous failures in f,,
v' License Event Reports,' by the establishment by the NRC of a l
Generic Task C-ll, Assessment of Failure and Reliability of f
~
Pumps and Valves, and by the ' establishment of a second Generic Task C-9, RHR Heat Exchanger Tube Failures.
No time was avail-able during MEB's work for the County to further particularize areas of concern at the Shoreham plant.
General concern exists due to the fact that this plant has been in the design mill for a substantial period of time and it is likely that older and P
12-6
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generally less well qualified components have been specified and procured.
j j
viii.
Condenser Tubes ~and Condensate' Demineralizers.
The t
concern here is almost identical to that expressed in Section t
vii above.
Unfortunat'ely, no specific particularization of equipment deficiencies have been possible due 'to lack of authorization to perform discovery, etc.
4.
REFERENCES 1.
NUEEG-0410, NRC Program for the ' Resolution of Generic Issues Related to Nuclear Power Plants.
2.
NUREG-0471, Generic Task Problem Descriptions, Category B, C, and D Tasks, (unissued).
3.
Risk Assessment Review Group Report,. US NRC (NUREG/
CR-0400).
5.
SUMMARY
AND CONCLUSION l
Numerous problems have been observed in operating reactors with regard to reliability and failure of various components and equipment.
The Shoreham plant is particularly suspect with i
regard to continuation of deficiencies as it represents an older
~
GE model line that has been in the design mill for some time.
The deficiencies identified in part 3, Basis For Contention 1
above, indicate areas where additional discovery work shculd 12-7
~ ~
j be performed or where suitable submittals should be put forward by Applicant and/or Staff if corrective action has
~
been taken at Shoreham.
Unfortunately, little specific investi-gation of detailed probrem. areas was performed by hHB prior to the tarmination-of MEB work for the' County, so particulari-
[
zation of these contentions has had to lie primarily on con-sultant's knowledge of the General. Electric design practices and ope' rating BWR performance.
This should not preditidice
~
the ' fact that these are areas of significant safety consequence in evaluating the future ' operation of the Shoreham facility.
A definitive finding of safety is required in these areas l
t before operation can be authorized.
i-
_)
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9 t
I i
t A
CONTENTION l'3a'. b, & c NUCLEAR SYSTEM MATERIAL FAILURES
.e 1.
STATEMENT OF CONTENTION 13a.
Intervenors contend that the Applicant and Regulatory Staff have not adequately demonstrated that the Shoreham nuclear system meets the requirements of 10 CFR Part 50, Appendix A, Criteria 14, 30, and 31 with regard to adequacy of material selection and control for the following systems and components :
1.
Resistance of feedwater spargers to in-service cracking.
ii.
Use of seamless 304 SS piping and control of carbon content in primary piping systems.
iii.
Use of nitrided stainless materials in control
]
drive internals.
iv.
Suitability of ' control rod drive collet cylinder tube material.
v.
Use of stainless materials in reactor internals.
vi Suitability of weld applied st.ainless steel for vessel internal cladding.
13b.
Intervenors further contend that the Applicant and Staff have not adequately demonstrated the effectiveness of the technology and methods available that are required to satisfy the inspection and tests specified by 10 CFR Part 50, Appendix '~
A, Criteria 32, 36, 39, and 45.
s..
13c.
Intervenors further contend that Applicant has not developed adequate' experimental data to verify the suita-bility of material selection design as required by 10 CFR Part 50, Appendix 3, Criteria III.
2.
CONTENTION CHRONOLOGY This contention, along with all other currently active County cententions, was originally described in County's Amended Petition to Intervene, dated September 16, 1977.
13-1 i
In the October 11, 1977 Prehearing Conference, the Board ruled that contention 13a, b, and c was acceptable and that discovery could begin.
This ruling was confirmed in the Board Order dated January 27, 1978.
3.
BASIS OF CONTENTION The basis of this contention in all. parts was well-defined in Coung's response to Applicant's Second Set of Interrogatories,
January 31, 1978.
This information is repeated here for purposes of completeness as is additional information now available.
Contention 13(a) Material Failures The wording of Contention 13(a), In County of Suffolk's Ammended Petition to Intervene dated September 16, 1977 should also have included Criterion 4 of 10 CFR 50 i
l Appendix A as well as Criteria 14, 30, and 31 listed..
i With that correction, the specific requirements of 10 CFR 50 Appendix A for the six issues listed in Contention 13(a)
I
~.
are as follows:
U
/ 1.
Compliance of the Shoreham feedwater spargers to Criterion 4 is questioned.
Criterion 4 states i
that " components important to safety shall be designed to accommodate the effects of and be compatible with l
the environmental. conditions associated with normal operation -- ".
11.
The use of seamless 304 stainless steel piping and the control of carbon content in the piping systems is f
deficient with regard to Criteria 4, 14, 30, and 31 l
I
l I
t for.the following reasons:
4 (1)
Criterion 4 requires that structures, systems, and components important to safety shall be designed to accommodate the effects of and be compatible with conditions associ-ated with normal operation --."
(2)
Criterion 14 requires that the " pressure boundaries shall be designed, fabricated and tested so as to have an extremely low proba-
. s) bility of abnormal. leakage, of rapidly propa-i l
gating failure and of gross rupture."
(3)
Criterion 30 requires that " components which l
are a part of the reactor coolant pressure boundaries shall be designed, fabricated, l
erecced, and tested to the highest qu'ality i
standards practical."
(4). Criterion 31 requires that "the reactor coolant pressure boundaries shall be de-
}
l signed --- the probability of rapidly propa-gating fracture is minimized.
The design shall reflect consideration of service
. temperatures and other conditions of the boundary material --- in deternining (1) material properties, - "
~
13-3
iii, iv, v.
Compliance of the material selection for these components with the requirements of criterion 4 with respect to environmental conditions cited in subparagraph i above is deficient.
In addition failure of these safety related components inside the reactor coolant pressure boundary could result in damage to' boundary materials themselves and jeopardize the system inte-grity, such that compliance with Criterion 14
"---reactor coolant pressure boundary shall be designed --- to have an extremely low probablity of abnormal leakage, cf' rapidly propagating failure,'and of gross rupture."
vi.
Same.as for subparagraph 11, above.
Resistance of the Shoreham feedwater spargers "to in-service
~
cracking is questionable because of the extremely poor experience of other BWRs of similar design over the past several years.
Details of crack failures observed at these similar reactors are documented in NUREG-0312, Interim b
Technical Report on BWR Feedwater and Control Rod Drive and Return Nozzle Cracking and, in the G.E. Report NEDO-21480 on the same subject.
Page 19 and 20 of NUREG-0312 provide a summary of BWR feed-water nozzle cracking problems and indicate that feedwater spargers have been replaced or repaired on fourteen 3Was 13-4
j j
1 i
that had been in service for three years or more.
Details of,
i 4
the specific failures in question are provided in the above
) ;
l-referenced reports.
Basically, the problem is believed to be one of a combination of environment and material fatigue.
The use of seamless 304 stainless steel piping at Shoreham is
(
j deficient in that it does not satisfy the requirement of
],
Criterion 4 in that "--- components important to safety shall
- j' be. designed to accommodate the effects of and be compatible j
with the environmental conditions associated with normal operation-- " nor does it meet the requirements of Criteria
']
l 14, 30, and 31 which require that "the reactor coolant l
]
pressure boundaries shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure and of gross rupture," and of control of quality of the as fabricated 4
condition and of the material to perform properly under 4
"---consideration of service temperatures and other con-i dicions of the boundary material under operating, maintenance, ]
testing, and postulated accident conditions--."
A gerieral statement of the problem can be found in the ACRS Generic Problems listing II B - 4, Stress Corrosion Cracking in BWR Piping.
This problem.is su=marired therein as follows:
"The Austenitic stainless steels are commonly used as piping material in many of the smaller 3WR lines.
A combination of weld sensitization, residual stresses, superposed loads, and oxygen equal to or greater than 0.2 ppm in the BWR coolant g
can lead to cracking, initiating on the inner surface and propagating through'the wall.
In mose cases there nill. be l
a leak well 'uefore pipe ' failure so there'is adequate warning; however, one can postualte 'a LOCA caused by. a guillotine
~
break with minimal prior warning.
Current ' efforts are to minimize stress' 'c'orrosion by using other materiah." (empha-sis added)
Details of these problems are found in NUREG-0313, Technical Report on Material Selection and Processing Guide-lines' for BWR Coolant Pressure Boundary Piping.
~
In NUREG-0410, NRC Program for the Resolution of Generic t
E Issues Related to Nuclear Power Plants, stress corrosion cracking in BWR piping is identified as' a problem to be
~
i corrected by a " policy" resolution, meaning that the resolution t
of the item is to be effected through administrative means l
rather than by a specific technical activity.
Policy reso-lution of the IGSCC problem has been attempted by the NRC in the past.
The NRC Standard Review Plan, NUREG-75/087, I
Section 5.2.3, Reactor Coolant Pressure Boundary Materials,
[)
states on page 5.2.3-3 that:
"Special provisions may apply to the use of austenitic stainless steel in Boiling Water Reactor (BWR) piping because plant operating experience indicates that reactor coolant boundary piping is susceptible to oxygen-assistad stress corrosion cracking."
In 1973, Regulatory Guide 1.44 ". effectively prohibit (ed) the i
use of welded-type 304 stainless steel by considering it the l
same as severely sensitized material and by placing limits on
(
oxygen concentration which cannot be met by operating BWR's."
i f
13-6 l
(238 Nuclear Steam Supply System GESSAR, page 5.2-22).
Subse-quantly, the NRC Staff's position was modified by the issuance of Branch Technical Position MTEB-5-7, which indicated that Regulatory Guide 'l.44 will be revised to provide ~ additional guidance on accep' table. practices.
Thus, it'would appear that the " policy resolution". attempted by the Staff, which " efface.
tively prohibited" the use 'of IGSCC-susceptibl materials, was defeated by a particularly teriacious reluctance to' change materials by General Electric.
Numerous occasions of piping failures have been experienced at BWR's all over the world.
The required shutdown of all B in the United States for pipe inspections in early 1975 is well known and initiated the ' formation of NRC and industry task forces to evaluate'the problem.
The NRC has recently (September,1978) reactivated the pipe crack cask force due to reports of IGSCC at BWR's in Ger.aany and Japan.
This cracking is of particular concern because it involves pipes as large as 24 inches in diameter, a size that had previously been
.3 considered to be of very low susceptibility.by the U.S. indu:.;jr.
Problems with such ' cracks are further complicated by the fact that ultrasonic inspection techniques are of questionable effectiveness.
Also of concern is the uncertainty of leak detectability versus crack growth as recently demonstrated at Duane Arnold.
All of these examples of serious material problems in BITR's,
coupled with the " older model" syndrome at Shoreham lead to 13-7
serious doubt that Shoreh'a'm in fact complies with 'the 6
applicable regulations.
The basis of Contention 13b, effectiveness of methods available to perform inspections and tests is also borne out l
Criterion 32 requires that "com-l by past and recent experience.
ponents which are a part of the reactor coolant pressure boundary shall be' designed to permit (1) periodic inspection f
and testing of important areas and features to. assess their structure and leak-tight integrity, and (2) an' appropriate I
surveillance program for the reactor pressure vessel."
The difficulty of complying with this requirement is well.su=marized in Task Action Plan A-14, Flaw Detection, contained in NUREG-0410, NRC Program for the Rasolution of Generic Issues.
The problem basically is one of developing effective non-destructive testing methods, equipment, and techniques usable in the restrictive environment constituted by a large, heavy walled reactor pressure vessel that is highly radioactive.
Task f
Action Plan A-14 currently has a scheduled completion date of
\\_.
Sep tember, 1980.
It is assumed that " technology and methods
---required to satisfy the inspection and tests specified" may be developed under this plan.
Technology, tests, methods, inspection and tests proposed to satisfy the inspection criteria covered by Criteria 36, 39 and 45 presu= ably will be provided in the Shoreham FSAR Technical Specification (Section 16).
This' document is not yet available but should be reviewed when available.
13-8 I
i The basis of Centention' 13c, Material Experi= ental Data, j
is contained in 10 CFR Part 50, Appendiz A Criterion 3 i
l which requires that "..... design control measures shali provide a
for verifying or checking the' adequacy of design such as by the i
performance of design reviews---calculational methods, or by the p.erformance of a suitable' testing program."
In view of l
1 the extensive in-service' failures reported in various documents such as NUREG-0410, NUREG-0312, NUREG-0313, and GE NEDO-21480, and because of the~ initiation ~ of extensive material testing programs by the General Electric Company, the Electric Power.- s
- )
i Research Institute, and others, it seems obvious that the
{
design control measures performed to date have been inadequate
)
and have not been successful in " verifying or checking the adequacy of design."
i In summary, it is readily apparent that serious safety issues result from the on-going material problems to be expected j
at the Shoreham plant.
General commitments by Applicant and Applicant's vendor to the use of alternate materials at some-
.. )
. time in the future without specific details.of when, where,
V and how at the Shoreham plant do norMng to assure resolution of these problems at this critical operating license stage.
4.
REFERENCES 1.
NUREG-0312, Interim Technical Report on SWR Feedwater and Control Rod Drive and Return No :le Cracking.
2.
NEDO-21480, General Electric Report, Interim Technical Reporr on SWR Feedwater and Control Rod Drive and Return No::le Cracking.
13-9
\\
3.
NUREG-0313, Technical Report..on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping.
4.
NUREG-0410, NRC Program for' the Resolution of Generic Issues Related to Nuclear Power Plants.
i 5.
NUREG-75/087, NRC Standard Review Plan, Section 5.2.3, Reactor Coolant Pres'sure Boundary Materials.
6.
238 Nuclear Steam Supply System GESSAR.
7.
Branch Technical Position MTEB-5-7.
8.
l' 9[
NRC Press Release, June 29,1978[
i
- 10. GE Reactor Pipe Cracks, CBE-7875, Oct. 6, 1978.
- 11. ACRS Subcommittee ' Meeting Transcript of September 22, 1977 meeting in San Jose, and September 23,.1977 meeting in Palo Alto, California.
- 12. Transcript of ACRS Subcommittee Meeting, September 11,12, 1978 in Washington, DC.
~ 13. NRC' Abnormal Occurrence Event, page 35134 -35135j Federal Register, August 8, 1978..
- 14. September 1978 Draft of NRC' Program Planned for Metallurgy and Materials Research Branch.
I
- 15. NRC Staff letter, September 28, 1978, regarding IGSCC experience in Germany and Japan.
- 5.
SUMMARY
AND CONCLUSIONS The problem of material failures is without a doubt one of the most serious issues facing the continued safe operation j
of BWR's in general, and Shoreham in particular.
Documentation of the problems and of the attempts to resolve thes are voluminous and could support days of investigation.
The fact 13-10 1
remains, however, that.the' problem has not been resolved on a generic basis and little specific *information has been provided by Applicant and/or Staff as.~to specific fixes o'r" programs intended at the Shoreham facility.'
It is essential that these problems be resolved to the gr'eatest e: stent possible before the Shoreham plant is permitted to start up so as to preclude future degradation of the safety margin at the plant and to avoid the unnecessary expenditure of personnel exposure as well as sub-stantial sums of money to correct these deficiencies after the plant becomes radioactive.
.'y e
9 e
8 e
O O
/%
W-G i
1 13-11
COhu.an0N 14a TURBDE MISSILES
'h 1.
STATEMENT OF CONTENTION 14a.
Intervenors contend that the Applicant has not adequately demonstrated that the Shoreham nuclear plant meets the require-ments of 10 CFR Part 50, Appendix A, Criteria 4, Environmental and Missile Design Bases, with regard to turbine orientation and/or missile shields.
2.
CONTENTION CHRONOLOGY This contention, along with all other currently active County contentions, was originally described in County's Amended Petition to Intervene, dated September 16, 1977.
In the October 11, 1977 Prehearing Conference the Board accepted the wording of Contention 14a and authorized the beginning of discovery.
Corifirmation of this ruling was found in Board's Order dated January 27, 1978.
3.
BASIS OF CONTENTION The. basis of this contention is found in 10 CFR Part 50, Appendix A, Criteria 4 which requires that " structures, systems,
and components important to safety shall be designed to accomodate the effects of...... postulated accidents....."
Such accidents have been interpreted to include gross failure of the turbine generator which could generate a missile that might damage reactor containment pressure boundary or other safety-related equipment.
The orientation of the Shoreham turbine generator A
14-1 l
,,i....
e-n-c--
7,_,.
with respect to the~. reactor is not optimmly designed to minimize the probability of such effects.
Regulatory Guide 1.115 specifies perpendicular orientation of the turbine generator center line to the' reactor location.
The tangential orientation found at Shoreham is in direct violation of this guide recommendation.
This problem is also included in NUREG-0410, NRC Program for the Resolution of Generic Issues, as Item A-37.
Task Action A-37 was issued in May 1978 and calls for a project completion of March 1979.
NRC's Task Action. Plan A.37 states on page A-37.1 that
" presently, there is no formal NRC guidance'on how to evaluate ^
the effect of toughness of turbine disc. material or tes, ting and inspection of discs, overspeed control systems, and steam valves on the missile generation probability.
Also, the acceptable methods for estimating missile strike probability for both low and high-trajectory missiles should be identified and documented.
Finally, a more thorough review is needed to evaluate the probability of damage that could lead to unacceptable radiological consequences."
This program, if completed on schedule in 1979, should provide additional information in time 'for inclusion in the Shoreham operating license hearing.
Other information likely to be available in this time frame is completion of the turbine missile tests being conducted at the Sandia Laboratories in Albuquerque, New Mexico.
I l
l 14-2
Since the Shoreham plant configuration obviously is not in compliance with Regulatory Guide 1.115, it is essential that
. these ~ critical reviews be completed and presented to provide additional guidance on. the 'a'eceptability of the Shoreham plant design.
Potentially vulnerable areas to turbine generated missile strikes may require additional shielding and/or pro-tection.
This is essential to complete the plant adequacy review since the FSAR indicates that " catastrophic failure of rotating equipment, namely pumps and turbines, flanged equip-
~
ment, and piping leading to the generation of missiles is not considered credible.'?
(SNPS-1 FSAR 3.5.2.u)
Applicant and Staff should submit documented results of appropriate missile analyses, with suif:able consideration of the incomplete Task 1
[
Action Plan A-37 to show that adequate consideration has been given to this accident possibility.
1 4.
REERENCES
~
1.
NUREG-0410, NRC Program for Resolution of Generic Issues Related to Nuclear Power Plants.
2.
Update of all Category A Task Accion Plans, presented as a part of NRC Staff testimony on ths Black Fox Construction Permit hearings October 1978.
o 3.
Regulatory Guidel.ll5, Revision 1, July 1977.
5.
SliMMARY AND CONCLUSIONS No definitive finding of safety can be made at this time regarding the Shoreham nuclear power station because 'of the 14-1
_ _7, 7. ;, _
inadequacy of missile review presented in the FSAR, and because
~
i of the incomplete ' nature of Task Action Plan A-37.
In addition, t
the Shoreham nuclear power plant configuration is not in com-pliance with Regulatory Guide 1.115 and it does not appear that appropriate consideration has been given to vulnerability of other critical safety systems and components to missile ' strikes.
It is assumed that suitable evaluation of this issue may yet be forthcoming by Applicant and Staff and that it will be accurately described in the SER.
This should be reviewed when the SER is issued and litigation of these contentions continu' d, if necess e
e 14-4 A--------.--
h i
CONTENTION 15a EMERGENCY RESPONSE PLANNING L
1.
STATEMENT OF CONTENTION 15a.
Intervenors contend that the Applicant and Regulatory Staff have not prepared an emergency response plan for Shore-ham in adequate compliance with the requirements of 10 CFR Part 50.34(b)'(6)(v) and 10 CFR Part 50, Appendix E including proper consideration of the adequacy of:
1.
Population evacuation area.
ii.
Federal, state, and local authorities' agreenents and plano.
iii.
On-site radiation treatment and decontamination facilities.
iv.
Off-site radiation treatment and decontamination l
facilities.
v.
On-site post accident monitoring equipment.
j vi.
Off-site post accident monitoring equipment.
vii.
Traffic monitoring on primary evacuation routes.
viii.
Plans for public training and " mock" evacuation drills.
b ix.
Instrumentation for automatic evacuation initia-l m
tion and notifiests.on.
j 1
- 2..COhn.a nON CHRONOLOGY t
This contention, along with all other currently active County. contentions, was originally described in County's i
Amended Petition to Intervene, dated September 16, 1977.
In accordance with the October 11, 1977 Prehearing Conference ruling, additional information was submicted by County in its November 10, 1977 Contentions and Legal Arguments document.on this contention.
The Board's order of January 27, 1973 15-1
~..
+
subsequently disallowed Contention 15a in total.
The County's l
February 17, 1978 objections to this order were filed and the Board's March 8,1978 order then found that Contention 15a 11 through vii and ix to be acceptable with the limitations that 15a vil is acceptable in so far as it relates to evacuation of the low population zone and 15a' ix is acceptable to the extent that it calls for instrumentation to initiate evacuation on the plant site or in the low population zone.
3.
BASIS FOR CONTENTION See County's Response to NRC' Staff's First Set of Inter-rogatories, dated August 23, 1978.
4.
REFERENCES See No. 3 above.
5.
SUMMARY
AND CONCLUSIONS County has been handling litigation of this contention without technical assistance from MI3.
Accordingly no infor-:.-
mation is included in this report.
e 9
15-2 n
~....
COhuanON 16a EVACUATION 6
i 1.
STATEMENT OF CONTENTION l
l 16a.
Intervenors contend that the Applicant and Regulatory Staff have not prepared and assessed an ade~quate evacuation plan for Shoreham as required by 10 CFR Part 50, Appendix E, Criteria IV-E and 10 CFR Part 50.34(b)(6)(v) with regard to the timing and feasibility of evacuation including, but not limited to, consieration of road networks, water barriers, weather conditions, and population density.
2.
CON ua nON CHRONOLOGY This contention, along with all other currently active County contentions, was originally described in County's Amended Petition to Intervene, dated September 16, 1977.
At the October 11, 1977 Prehearing Conference, this contention was put in the category of those to be held abeyance for a decision subject to further discussion and argument.
Such information was submitted by Cotnty in' their November 10, 1977 document covering amended contentions and legal arguments.
The Board's January 27, 1978 order disallowed this contention.
However, subsequent to County's objections to the Board's i
January 27, 1978 order, the Board in its March 8,1978 ruling found that Contention 16a should be admitted to the e' tent that x
the contention relates to evacuation within a low population cone.
i 3.
BASIS FOR CONTENTION See County's Response to NRC S taff's First Set of Inter-rogatories, dated August 23, 1978.
l 16-1
l 4.
REn.ar ACES See No. 3 above.
l i
5.
SUMMARY
AND CONCLUSIONS County has been handling litigation of this contention without technical. assistance from NHB.
Accordingly no infor -
mation is included in this report.
r%
E e
16-2
CON uan0N 17a STlT SUITABILITY s
n 1.
STATEMENT OF CONTENTION 17a.
(9/16/77 wording)
Intervenors contend that the Appli-cant and the Regulatory Staff did not adequately review the site for Shoreham for compliance with 10 CFR Part 50, Appendix A, Criteria I and 10 CFR Part 100 with regard to:
1.
Physical characteristics of the site including G-value selected for the Safe Shutdown Earthquake and Operating Basis Earthquake.
ii.
Determination of exclusion area, low population zone, and population center distance.
111.
Feasibility of receiving and shipping radioactive material.
iv.
Feasibility of implementing an evacuation plan for plant employees and the public.
17a.
(11/10/77 wording)
Intervenor contends that the Appli-cant and the Regulatory Staff did not adequately review the site for Shoreham for compliance with 10 CFR Part 50, Appendix A, Criterion 2 and 10 CFR Part 100 with regard to its seismic characteristics, including the adequacy of the ground accelera-tion value selected for the Safe Shutdown Earthquake (SSE) and the Operating Base Earthquake (OBE),
i 2.
CONTENTION CHRONOLOGY This contention, along with all other currently active County contentions, wa~s originally described in County's Amended Petition to Intervene, dated September 16, 1977.
In the October 11, 1977 Prehearing Conference, the Board requested additional information on this contention and the wording of the contention as presented above was revised in County's November 10, 1977 Amended Contentions and Legal Arguments document.
In the Board's January 27, 1978 ruling, this conten-tion was.' disallowed.
Following the County's Feb uary 17, 1973 17-1
objection to Board's January 27 order, the Board issued a March 8,1978 order which ruled 'that Contention 17a, parts i and 11 (per the September 16, 1977 version) were acceptable for discovery but must be later refined.
3.
BASIS OF CONTENTION The basis for Contention 17a(i), characteristics of the site including acceleration value for SSE and OBE, was pre-i sented in County's Amended Contentions and Legal Arguments submitted on November 10, 1977 and was in part repeated in i
County's Objections to Hearing Board's January 27, 1978 Memorandum and Order, submitted February 17, 1978.- The basis of concern of earthquake design bases is primarily due to the current reassessment of " eastern" seismicity.
An example of this reassessment found in NRC B. Rushche's January 1977' letter to ACRS asking for cletrification of such bases.
This request was presumably brought about by seismologists' concerns over lack of data base necessary to predict earthquakes in the.
eastern part of the United States and by reports of new infor../
mation concerning the ' seismicity of the New England, New York, and New Jersey coastal areas.
(See Science, April 28, 1978,
" Earthquakes, Faults, and Nuclear ?cwer Plants in Southern New York and Northern New Jersey',' by Yash P. Aggarwal and Lynn R.
Sykes.)
The applicability of such concern and uncertainty specifi-cally applies to the Shoreham plant and is reinforced by the NRC's K. Kneil's September 1,1977 letter to the Applicant 17-2
I addressing the open items of seismic qualification of 4
mechanical equipment at Shoreham, electrical equipment, k
and load combinations, including effect on concrete con-tainment structures.
No significant amount of investigation on seismic ade-quacy of the site has been performed for the County by MHB other than as seisnic loading pertains to load combination questions with other contentions.
However, it is readily apparent that a significant question exists as to the ability to predict seismic events in regions where such events occur infrequently.
NRC deliberation on the whole question of seismic design basis for eastern plants is producing a pressure for higher design values.
In the case of Shoreham, this is particularly significant with regard to containment LOCA and SRV load evaluations, since less conservative methods of load assessment are proposed, making it essential that non-conservatisms not be introduced by seismic evaluations.
The~ basis for Contention 17a(ii), determination of exclu-
,1 sion area, etc., is touched upon in the discussion on Contentions 6a and b.
If the site seismic design basis is underassessed, a seismic event could result in accidents with consequences greater than the design basis accident, with associated detri-mental effects on surrounding populations.
This, in effect, makes the size of the LPZ and exclusion area an issue of contro-i versy.
17-3
i 1
j 4.
REFERENCES 1.
Science, April 28, 1978, " Earthquakes, Faults, and Nuclear Power Plants in Southern New York and Northern New Jersey, Yash P. Aggarwal and Lynn R. Sykes.
2.
Letter, B. Rusche to ACRS, January 1977.
3.'
Letter, NRC Kneil to LILCO, September 1,1977.
4.
NUREG-0410, Update Task Action Plan A-40, Seismic Design Criteria.
5
SUMMARY
AND CONCLUSIONS In view of the uncertainty existing in assessment of N
seismic design values at Shoreham, it is recommended that thit
,1 l
contention be followed closely in conjunction with Contentions 9 and 10, Containment Adequacy.
Seismic, LOCA, and SRV tran-sient load combination methodology must be considered as an interrelated problem.
Applicant, and in particular NRC Staff, should address the as yet unresolved question of reconsideration of seismic values for eastern U.S. plant locations.
Impact of this reassessment should be factored into determination of
-h Shoreham exclusion area.
It should also be determined if additional steps should be taken in evacuation planning.
17-4 e.
l CONTENTION 18a 3
SABOTAGE 4
4 i
1.
STATEMENT OF CONTENTION 18a.
Intervenors contend that the Applicant has not developed an adequate security plan for Shoreham, as required by the current requirements of 10 CFR Part 73.50 and 10 CFR Part 50.34(b)(8c), for guarding against domestic sabotage of the facility, including proper consideration of:
1.
Intentional airplane crashes.
ii.
Boinbs.
iii.
Hij acking.
iv.
Blackmail.
v.
Paramilitary attacks.
vi.
Terrorism.
vii.
Theft of nuclear material.
viii.
Short and long term health effects.
ix.
Effectiveness of countermeasures.
x.
Automatic shudown of vital equipment.
xi.
Communication to off-site law enforcement centers.
I 2.
CONTENTION CHRONOLOGY This contention, along with all other currently ' active County contentions, was originally described in Cotmty's l
Amended Petition to Intervene, dated September 16, 1977.
In accordance with the decision made at the October 11, 1977 Prehearing Conference, additional information was submitted on.this contention by County with its November 10, 1977 docu-
=ent on Amended Contentions and Legal Arguments.
Subsequently, 18-1
the Board ruled on January 27, 1978 that this contention was acceptable for discovery on parts 11, v, 'vi, vii, ix, and xi.
}
-, q i '
' j 3.
BASIS FOR CONTENTION See County's Response to NRC Staff's First Set of Inter-
- 1 i
rogatories, dated August 23, 1978.
4.
REFERENCES See No. 3 above.
f, 5.
SUMMARY
AND CONCLUSIONS N
\\
J County has been handling litigation of this contention without technical assistance from MHB.
Accordingly, no infoi-mation is included in this report I
l l
l 4
18-2
j CONTENTION 19a f
FUEI;/ WASTE TRANSPORTATION t
1.
STATEMENT OF CONTENTION 19a.
Intervenors contend that the Applicant and Regulatory Staff have not adequately demonstrated that the transportation of fuel and radioactive wastes to and from the Shoreham site will comply with 10 CFR Part 71, with respect to:
1.
Compliance with New York City codes and regu-lations including Article 175 entitled, "Radia-tion Control," by the New York City Health Department.
s 11.
Radiation exposure in New York City's densely populated metropolitan areas as a result of a routine and accidental release of spent fuel and/or wastes during transportation.
iii.
Risk due to rupture of a spent fuel cask, including prompt and latent fatalities and illnesses, p'lus the cost of decontamination measures.
iv.
Meteorological conditions which prevail along the proposed transportation routes.
v.
Security for in-transit nuclear materials, including possibility of sabotage and/or diversion.
vi.
Health hazards analysis of barge shipments of nuclear materials through Long Island Sound.'.
2.
CONTENTION CHRONOLOGY This contention, along with all other currently active County contentions, was originally described in County's Amended Petition to Intervene, dated September 16, 1977.
In accordance with the October 11, 1977 Prehearing Conference.
ruling, additional information on this proposed contention was submitted by County with their November 10, 1977 document on Amended Contentions and Legal Argu=ents.
The Board's l
19-1
January 27, 1978 order ruled that this contention, as worded, was beyond the scopc. of the operating license hearing.
County's objection to this ruling, dated February 17, 1978, provided additional argument on contention validity, but Board's March 8,1978 order ruled that this contention was unacceptable as a challenge to regulations.
3.
BASIS FOR CONTENTION S'ee' County's objections to Board's January 27 order document dated February 17, 1978, pages 18 through 20.
4.
REFERENCES See Item 3 above.
5.
SUMMARY
AND CONCLUSION No technical advise is being provided County by MIB on this contention.
Since it has been ruled beyond scope of the heariag, it will be given no further discussion in this report.
l 1
19-2
CONTENTION 20a SYSTEM DESIGN FOR ACHIEVEMENT OF "ALARA" RADIATION EXPOSURES
'4 1.
STATEMENT OF CONTENTION 20a.
Intervenors contend that the Applicant has not adequately demonstrated that the Shoreham nuclear system meets the require-ments of 10 CFR Part 20.l(c), Standards for Protection Against P.adiation with regard to provisions relevant to maintaining occupational radiation exposure as low as is reasonably achievable (ALARA).
Demonstration of compliance is inadequate -
in the areas of:
1.
Plant and equipment design has not been shown to be optimumly developed for minimization of radiation exposure during maintenance of the plant by:
(1)
Selection of low cobalt materials.
(2)
Separation or isolation of various components and piping systems.
(3)
Provisions for flushing or decontami-nation.
(4)
Equipment layout and arrangement for ease and automation of maintenance and refueling.
ii.
Provisions in the' system design to facilitate future plant decommissioning.
2.
CONTENTION CHRONOLOGY q,
This contention, along with all other currently active County contentions, was originally described in County's Amended Petition to Intervene, dated September 16, 1977.
The Board ruling in the October 11, 1977 Prehearing Conference was that this contention in total was found acceptable.
This decision was confirmed in the January 27, 1978 Board order, although the footnote on page 2 of that order referred to 20a (vi) miscakenly instead'of 20a (ii).
This error was corrected 20-1
in the Board's March 8,1978 order and the contention remains acceptable as originally stated.
3.
BASIS FOR CONuan0N The regulatory basis for Contention 20a is that. the Appli-q cant has not adequately demonstrated that the Shoreham plant meets requirements of 10 CFR Part 20.l(c), Standards for Pro-taction Against. Radiation, with regard to assuring that the design complies with the requirements of As Low As Reasonably Achievable (ALARA) guidelines.
No mention is made in the FSAR of Regulatory Guide 8.8, Information Relevant to Ensuring tha".)
Occupational Radiation Exposures at Nucle ~ar Power Stations Wi.1.f be As Low As is Reasonably Achievable, and Section 12.I of the FSAR on Radiation Protection discusses "As Law as Practicable" (ALAP) instead of ALARA.
FSAR Section 12.1 focusses almost exclusively on procedural control of occupational radiaticn expo'sure.
Some discussion of features required to mitigate operating radiation levels are claimed as ALAP considerations, but little recognition is given to the fact that currently operating nuclear power plants are' finding that 757. or more of. annual occupational radiation expo-sure is accumulated during periods of plant shutdown.
No consideration appears to have been given to the selection of materials to minimize the buildup of radioactive crud such as the selection of low cobalt piping alloys or avoidance of the use stellite in valve trim to reduce cobalt-60 levels.
20-2 l
...t For example, FSAR Part 12.1.2, Design Considerations, indicates that valve selections are made based on best product available I
considerations with no apparent thought given to the minimzation
[
I of generation of activated corrosion products, ease of main-
[
i l
tenance, or minimization of crud pockets.
Credit is taken for i
j radioactive piping being located behind shielding to minimize
[
radiation exposure to operating personnel, but no mention is j
made of the fact all drywell piping, including recirculat2.on L
t pumps, isolation valves, mainsteam isolation valves, and safety-
[
relief valves are located in a common space where personnel required for maintenance and inspection operations will be i
required' to be expo' sed radiation doses generated by all co:: mon f
equipment in that space.
No mention is made of provision for i
flushing or decontamination of primary piping and equipment, although credit is taken for providing auxiliary systems with j
pipe flushing connections.
Additionally, no mention is made of provisions in the system design to facilitate future plant I
decommissioning.
In fact, the monolithic concrete construction l
of the Mark II type containment provides a difficult decom-L r
i missioning task.
i t
Additional deficiencies in ALARA provisions undoubtedly exist.
An example of such deficiencies is contained in NUREG-l l
0312, Interim Technical Report on BWR Feedwater and Control Rod Drive Return Line No::le Cracking.
It is stated on page 12 of this repert that the inspection and repair of reactor vessel no cle cracks has potential for significant occupational radia-tion exposure because of the high radiation levels in the work L
20-3 l'
= -
_-.7.,
k areas and relatively long stay time required to perform the necessary work.
The fact that nuclear plants are being licensed with known generic problems such as nozzle cracking potential, deficiencies in material selection and unreliable ti valves and equipment by definition is a violation of ALARA principle.
The problem of maintaining occupational radiation exposure in compliance with ALARA requirements is just beginning to be 4
addressed seriously by the industry in general and Shoreham, j
being an older version of the BWR has particular problems.
A i
presentation by the NRC Staff to' the ACRS on October 6,1978 j
reveals that the NRC was in the process of upgrading occupational radiation exposure from a Category 3 generic problem to a problem i
j of highest priority.
Decommissioning of reactor plants at end of life is also under active consideration for upgrading.
These
~
changes are undoubtedly being considered in view of the current f-mood to substantially reduce allowable occupational exposure i_
levels by as much as a factor of 10.
Such reductions will un-1 doubtedly occur if not. i=diately within the relatively near future, and it is essential that all nuclear plants be thoroughly reviewed for compliance with ALARA guidelines as early as possible.
This should preferably be done at the construction permit review but a thorough review at the operating permit stage is vastly superior to attempting to make corrections after the plant becomes radioactive.
20-4
.s 4.
REFERENCES 1.
ACRS Full Committee Meeting. transcript, Oct. 6, 1978.
't 2
ACRS Environmental Subcommittee Meeting Transcript, Jan.
25-26, 1978.
3.
Regulatory Guide 8.8, Information Relevani: to Ensuring That Occupational Radiation Exposure at Nuclear Power Stations Will Be As Low As is Reasonably Achievable, l
Revision 2, Mar. 1977.
4.
NUREG-0323, Occupational Radistion Exposure at Light l
Water Cooled Power Reactors, 1976.
5.
NUREG-0471, Generic Task Problem Descriptions, Category I
B, C, and D Tasks, Unissued Report presented as a part of NRC Staff testimony on generic issues at the Black Foc Construction Permit Hearings.
6.
NUREG-0410, NRC Program for the Resolution of Generic Issues Related to Nuclear Power' Plants.
7.
An Overview of the Activated Corrosion Product Reduction-I Program for U.S. Power Reactors, T.D. Murphy, Radiation Protection Section, NRC.
8.
Getting at the Source; Reducing Radiation Fields, Robert
~
A. Shaw, EPPl.
9.
Bulletin of Atomic Scientist,: Sep.1978, " Cancer and Low Level Ionizing Radiation," Karl Z., Morgan.
10.
Bulletin of Atomic Scientist, Sep.1978, "The Risks for Radiation Workers." J. Rotblat.
L 5.
SUMMARY
AND CONCLUSIONS In reviewing the Shoreham FSAR and submitted documentation, it rapidly becomes obvious that minimization of occupational radiation exposures and compliance with ALARA requirements is at best an exercise of rationalization and afterthought.
Cem-pliance with the requirements is claimed as a result of promises l
to provide procedural control in order to overcome the short-ccmings of material selection, equipment layout and lack of shielding for maintenance protection.
If the NRC is serious 4
20-5
o.s about resolving the generic issue of occupational radiation exposure, *: hey will require a complete and thorough evaluation of this problem on the Shoreham plant and will insist on
~
I hardware changes where possible.
i Respectfully ~ submitted, ht /8d [*)
Irving Like.
Special Counsel for the County of Suffolk Dated:
November.30, 1978 h
s e
e W
e 20-6
~,t UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC S *..NGs i t% i GENSING SOARD CV N
.3 Ge s 3
x$
.,o 7 k'-
4 ;-
49,$ CT 6
In the Matter of 1
j)
%d 51
~T' g++?j9 LONG ISLAND LIGHTING COMPANY 9
'J eket No. 50-322 s
) 9 y
(Shoreham Nuclear. Power Station #
Unit 1) f/msjtd\\
CERTIFICATE OF SERVICE I hereby certify that on Nov. 30,1978
, a copy of COUh""Y OF SUFFOLK'S PARTICULARIIED COICENTIONS was sent by postage pre-paid, first class mail, to each of the following:
Elizabeth S. Bowers, Esq.
Edward M. Barrett, Esq.
1 Chairman General Counsel Atomic Safety & Licensing Board Long Island Lighting Compa.my U.S. Nuclear Regulatory Comm.
250 Old Country acad Washington, D.C.
20555 Mineola, New York 11501 Dr. Oscar Paris, Member Edward J. Walsh, Esc.
Atomic Safety & Licensing Board Long Island Lightins Company U.S. Nuclear Regulatory Comm.
250 Old Country Road Washington, D.C.'
20555 Mineola, New York 11501 Frederick J. Shon, Member W
Taylor Reveley, III, Esq.
Atomic Safety & Licensing Board Hunton & Williams U.S. Nuclear Regulatory Com=.
P.O. Box 1535 Washington, D.C.
20535 Richmond, Virginia 23212 Docketing and Service Section Jeffrey C. Cohen, Esc.
k Office of the Secretary N.Y. State Energy Office U.S. Nuclear Regulatory Comm.
Swan Street Sldg.
Core 1 Washington, D.C.
20555 E=pire State Plaza Atomic Safety & Licensing Appeals Scard Howard L. Slau, Esq.
U.S. Nuclear Regulatory Cc=m.
Slau & Cohn, ?.C.
Washington, D.C.
20555 380 North Broadway Jericho, New York 11753 Richard Hoefling, Esq.
Atomic Safety & Licensing Scard T.J.
Surke U.S. Nuclear Regulatory 00:m.
Project '4anager Washington, D.C.
20555 Shoreham Nuclear ?:ver 5:a:i:r F.O.
Ecx 613
'E3 Technical Associates North Ocuntry Road 366 California Ave. - Suite 6 Wading River, ::ew Y:r!:
11722 Palo A1:c, California 94306
't j
f I k G'.
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