ML20059H128
| ML20059H128 | |
| Person / Time | |
|---|---|
| Issue date: | 09/05/1990 |
| From: | Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Cranston A SENATE |
| Shared Package | |
| ML20059H132 | List: |
| References | |
| DD-89-09, DD-89-9, NUDOCS 9009170003 | |
| Download: ML20059H128 (3) | |
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W NUCLEAR REGULATORY COMMISSION.
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September 5, 1990 s
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The Horrrable Aler. Cranston United,tates Senate Washington, D.C.
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Dear Senator Cranston:
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1 atri responding to your letter of August 13, 1990, which enclosed correspondence of May 12, 1990, from Ms. Wanda S. Fallentine of Mill Valley, Celifornia. You asked us to look into Ms. Ballentine's concerns with resuct to boilitig-water reactors tesigned and nanufactured by the General Electric Corr.pany.
Ms. Bellentine's May 12, 1990 letter expressed concerns regarding the continutd-operation cf all nuclear power reactors designed by the General Electric t
Con pany. As the basis for this concern, Ms. Ballentine 611estd that (1) in ofreactorsGeneralElectricbuildsbebanned;(2))in1986,anNRCofficial 1972, the U. S. Nuclear RtCulatory Commiissicn (NRC recope+nded that the type
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I adniitted that General Electric cot. taint: tnt vassels have a 90 percent chance of' failure in a nuclear accident;1 and(3)in1975,the"ReedReport,".(Erclosure3) i written by General Electric engineers, severely criticited the reactors and was
'l Ittt secret by General Eltctric and the NRC until they were forced to release it in 1987.
t InDirector'sDecision(DD-89-9),ofDecember4,1989,werespondedtoMs. Anna Harlowe's March 8, ISE9 letter that conveyed concerns siniilar to those expressed by.Ms. Baller. tine.
Briefly, in our Director's Decision.we concluded that General Electric reactors are currently designed to respond adequately.to all
- design-basis accidents, and that several NRC programs atd industry initiatives att being impleniented to in. prove the design 60d perfornience of General _ Electric reactors. After careful study, we determined that continued operation of these reacters is safe and does not require that the plants be shut down while improvements are being e.ade. Currently, our Director's Decision represents the NRC's resolve on the safe operation of General Electric reactors.
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'The Honorable Alan Cranston l Enclosed, as requested, is Ms. Ballentine's May 12, l'300 letter and our Director's Decision (00-89-9)..I trust this reply responds to your request.
If I can be of any further assistance, please contact me.
Sincerely, Melnsi Signed g,
y James M. Tayloy, James M. Taylor Executive Director for Operations
Enclosures:
1.-
May 12, 1990 Letter to Senator Alan Cranston from Ms. Wanda S.
Ballentine 2.
Director's Decision (DD-89-9) dated December 4. 1989
- 3.
NUREG 1285, " Reed Report
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Director {sDecision(DD-89-9).- I trust this (ply resperids to your r quest.
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If I can be of any furthtr assister.ce, please contact i,e.
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James M. Taylt/
h Executive D)r'ector for Operations
Enclosures:
/
- 1..May 12, 1990 Letter to Senator Alan Cianston frtim Ms. Wande S.
Ballt-ntine 2.
Director'sDecision(00-89-9) dated /
December 4, 1989
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- Reed Report"
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i 00-89 9 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMIS$10N 0FFICE OF NUCLEAR REACTOR REGULAT!0N Thomas E. Murley, Director.
In the Matter of BOSTON EDISON CO. (Pilgrim Nuclear Power Station, Docket No. 50 293)
{
CAROLINA POWER 8 LIGNY C0. (Brunswick Steam Electric Plant, Units 1 and 2,
)
Docket Nos. 50-324 and 50-325) l CLEVELAND ELECTRIC ILLUMINATING C0., ET AL. (Perry Nuclear Power Plant, Unit 1, l
Docket.No.50-440)
COMMONWEALTHEDISONCO.(DresdenNuclearPowerStation, Units 2and3. Docket Nos. 50-237 and 50-249), (Quad Cities Station, Units 1 and 2. Docket Nos. 50-254 and 50 265), LaSalle County Station, Units 1 and 2 Docket Nos. 50-373 and 50-374) t CONSUMERS POWER CO. (Big Rock Point Nuclear Plant, Docket No. 50-155) t DETROIT EDISON CO. (Enrico Femi Atomic Power Plant, Unit 2. Docket 'No. 50-341)
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GENERAL PUBLIC UTILITIES (0yster Creek Nuclear Power Plant Docket No. 50-219)
L GEORGIA POWER CD. (Edwin 1. Hatch Nuclear Plant, Units 1 and 2, Docket Nos.
50-321and50366)
GULF STATES UTILITIES C0. (River Bend Station, Docket No. 50-458)-
ILLIN0!$POWERCO.(ClintonPowerStation,DocketNo.50-461) n 10WA ELECTRIC LIGHT & POWER CO. (Duane Arnold Energy Center, Docket No. 50-331)
LONG ISLAND LIGHTING CO.- (shoreham Nuclear Power Station Docket No. 50-322)
MIS $1SSIPPI POWER & LIGHT CD. (Grand Gulf Nuclear Station, Dccket No. 50-416)
NEBRASKA PUBLIC POWER DISTRICT (Cooper Nuclear Station, Docket No. 50-298)
N!AGARA M0 HAWK POWER CORP. (Nine Mile Point Nuclear Station, Units 1 and 2 DocketNos.50-220and50-410)
NORTHEAST UTILITIES (Millstone Nuc1e' ar Power Station, Docket No. 50-245)
NORTHERN STATES POWER CO. (Monticello Nuclear Generating Plant, Docket No. 50-263) n : l'!W DTW R
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- ENNSYLVANIA POWER & LIGHT CD. (Susquehanna Steam Electric Station, Units 1 and2.DocketMos.50-387and50-388)
PHILADELPHIA ELECTRIC CD. (Peach Bottom Atomic Power Station, Units 2 &nd 3 Docket Nos. $0-277 and 50-278), (Limerick Generating Station, Unit 1, DocketNo.70-352)
POWERAUTHORITYOFTHESTATEOFNEWYORK(JamesA.FitzpatrickNuclear Power Plant, Docket No. 50-333)
PUBLIC $ERY!Ct ELECTRIC 8 GAS CO. (Hope Creek Nuclear Station, 0:,?aet No.
50-354)
TENNESSEE VALLEY AUlHORIT' (Browns Ferry Nuclear Power Station, Units 1, 2, and 3. Docket Nos. 50-PG, 50-260, and 50-296)
VERMONTYANKEENUCLEARPOWERCORP.(VermontYankeeNuclearPowerStation, DocketNo.50-271)-
WASHINGTO' PUBLIC POWER SUPPLY $YSTEM (WNP Unit 2, Docket No. 50-397)
DI' RECTOR'S DEcl$10N UNDER 10 CFR 2.206 INTRODUCTION On March 8, 1989, Ms. Anna Harlowe, on behalf of the Ecology Center of SouthernCalifornia(Petitioner),filedaPetitioninaccordancewith10CFR 2.206 with the Nuclear Regulatory Comission (NRC). The Petition was referredtotheDirector,OfficeofNuclearReactorRegulation(NRR),for consideration.
The Petition asked the Director, NRR, to fix or close all nuclear reactors designed by the General Electric Company (GE). As a basis for this request, the Petitioner alleged the following:
(1) In 1972, a member of the NRC staff recomended that GE-designed reactorsbebannedintheUnitedStates;(2)in1975,SEengineersgenerated the " Reed Report" that detailed dozens of safety and economic problems with I
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GE-designed reactors and recommended that GE stop selling those reactors; j
(3) in 1986, an NRC official admitted that 24 GE reactors with Mark I i
containments had a 90 percent chance of failure in a nuclear accident; (4) in 1987, an NRC task force confirmed that Mark I containments were virtually-ce'rtain to fail in an accident; (5) according to NRC safety studies, Mark !!
reactors have many possible scenarios for early containment failures; and (6) Mark !! designs, on which the Reed Report focused, have dozens of safety and economic problems and have suffered massive cost overruns during construc-
. tion as a result of design problems. Ms. Harlowe also expressed concern that the GE Advanced Boiling Water Reactor design " fails to address many of tho'
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shortcomings icentified by General Electric's own engineers as far back as the e
1975ReedReport"(Petition,p.2).
On June 5, 1989, ! scknowledged receipt of the Petition.
I informed Ms. Harlowe that (1) the Petition would be treated under 10 CFR 2.206 of L
the Commissior.'s regulations, and (2) appropriate action would be taken 1
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within a reasonable amount of time.
For reasons discussed below, the Petition is denied.
- 11. BACKGROUND The Petitioner alleges that in 1972, a Nuclear Regulatory Comission t
staff member recommended that GE-type reactors be banned in the United States.
It appears that the Petitioner is making reference to a memorandum by Dr. Steven Hanauer dated September 20, 1972. Specifically, Dr. Hanauer was concerned that then recently highlighted safety disadvantages of pressure-suppression containments might outweigh the safety advantages. He recommended that the Atomic Energy Comission (predecessor to the Nuclear Regulatory 1
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Connission) adopt a policy to discourage further use of pressure suppression.
containments and that such designs not be accepted for construction permits filed 2 years after the policy would be adopted.
The Petitioner also refers to a 1975 GE document known as the " Reed i
Report." The Reed Report was a self-critical study performed oy GE staff in 1975.
It was intended as a product improvement study to enhance the avail-abilityandperformanceofGE'sboilingwaterreactors(BWRs).
The report, I
by its nature a candid self-analysis, was intended for GE's internal use only.
It had always been held by GE to be ' proprietary," and thus not subject to public disclosure. The principal author of the report was j
Dr. Charles E. Reed, a Senior Vice President of GE. Contributors included-
' technical and professional personnel from a va'riety of GE departments.
Their efforts resulted in the Nuclear Reactor Study, referred to today as the Reed Report, and a set of 10 subtask reports that provided the detailed technical information used to develop the Nuclear Reactor Study.
L The Reed Report addressed operating BWRs and the design of future I
GE products and services-in the nuclear field. For reactors in operation l
at the time, the report discussed ways to improve a plant's availability and its electrical generating capacity factor through improvements in plant hardware and also in service, fuel, equipment, and operating procedures.
For future reactors, the report considered GE's then-new BWR design, the, BWR-6, and discussed problems regarding final design details, licensing, and full-power operation of BWR-6 plants.
The Petitioner also refers to an early 1986 statement by a senior NRC official that the containment vessels on 24 GE reactors have a 90 percent i
chance of failure in a nuclear accident. Ms. Harlowe most likely is referring V
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to a quote from Harold Denton in Inside NFj, Vol. 8, No.12 June 9,1986, wherein Mr. Denton was quoted as saying:
'I don't have the same warm feeling
'about GE containment that I do about the larger dry containee.nts. There has been a lot of work done on those containments,'but Mark I containments, especially being smaller with lower design pressure - and iri spite of the 1
suppression pool - if you look (at the) WASH 1400 reg safety study, you'll find something like a 901 probability of that containment failing."
The Petitioner also alleges that a late 1987 finding of an NRC task
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force confirmed that the failure rate of these 24 Mark I reactors is such that their containments are " virtually certain" to fail in an accident.
Although it is not clear which specific study the Petitioner is referring to, it f
is presuned that sha refers to the " Reactor Risk Reference Document," Draft NUREG-1150, dated February 1987. NUREG-1150 estimated the probability of total core damage frequency for the Peach Bottom reactor, which is similar in cesign to the typical Mark I reactor, to be 8.2 X 10-6 per reactor year..
However, NUREG-1150 went further and evaluated Mark I and other reactor design' risk scenarios given that a s'evere (core-melt) accident (low proi-ability event) had already taken place. Accounting for connents received from the public and three formal peer reviews, a second draft for peer review titled " Severe Accident Risks: AnAssessmentforFIveU.S. Nuclear Power Plants. Summary Report, Second Draf t for Peer Review " NUREG-1150, was issued in June 1989 in two volumes. Volur< 1 provides summaries of the risk analysis results for the five plants studied, perspectives on these results,
't and a discustion of the role of these risk analyses in the NRC staff's i
severe accident regulatory program. Volume 2 provides a more detailed
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discussion c# the methods used in the risk analyses, additional discussion I
on specific technical issues important in the analyses, and responses to connents received on the earlier draft.
I Petitioner also alleges that Mark !! reactors (eight of which are
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operating) still have many possible scenarios for early containment failure
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according to NRC safety studies.
Petitioner is most likely referring to studies conducted as part of the Containment Performance Improvement, Individual Plant Examinations, and Severe Accident Policy programs. NRC l
. studies are ongoing and not yet complete, but the NRC has made preliminary L
specific assessments of Mark !! containment performance.
Lastly, Petitioner alleges that " Mark !! reactors on which the 1975 General Electric Reed Report was primarily focused have the aforementioned l
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' dozens of safety and economic problems,' and have suffered massive cost f
overruns during construction as a-result of design problems." It is believed, based on the staff's review of the Reed Report, that Petitioner is referring to Mark !!! reactors, not Mark !! reactors, and it is on this premise that ny discussion is based.
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A.
Mark I Containment Concerns Petitioner's alleged " facts" that she wishes placed under consideration for relief contain three items that appear to be, directed at the GE Mark !
containment design. These are (1) that "in 1972 a Federal Nuclear Regulatory t
Comission [ sic) staff member recommended that General Electric-type reactors be banned in the United States," (2) that in 1986, "a top Nuclear Regulatory Comission official admitted that the containment vessels, the last barrier to
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radiation release, on M GE reactors have a 90.oercent chance of failure in-a nuclear accident," and (3) that "in late 1987, a Nuclear Regulatory Commission. task force confirmed the failure rate of these 24 ' mark I' reactors, saying that their containments are vjrtually certain to feil in an acc1 dent'." U l
Petitioner coes not provide any information of which the staff was i
unaware.- In fact, similar, more specific'end detailed concerns relative to alle ed Mark I containment design deficiencies were previously addressed in Interim Director's Decision 87-14 concerning the Pilgrim Nuclear Power Plant on August 21, 1907. U As stated in that Decision, containment structures are an integral part of the U.S. reactor designs in, that they form one part of a' structured, tiered approach to public safety known as defense in depth.
I Concisely put, cafense in depth is the process implemented by the AEC (later
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t NRC) to ensure that multiple levels of assurance and safety exist to'Mnimize the risk to the public of exposure to ionizing radiation resulting from t
equipment failures, transients and postulated accidents.
i A primary level of assurance are those activities to ensure that the plant l
~ is designed and constructed to high quality standards. The Commission's regula-tions require plant design to satisfy certain standards, as speci.fied in the General Design Criteria (GDC) in 10 CFR Part 50 Appendix A.
Specific information is provided in the NRC's Standard Review Plan (SRP) which details acc6ptable methods for complying with the requirements established in the GDC.
MEcologyCenterofSouthernCaliforniaPetitionat1.
M Boston-Edison Co. (Pilgrim Nuclear Generating Station) DD-87-14, 27 K'E O 7 (1987).
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j Early in the development of commercial nuclear power, it was recognized i
j that these complex systems could not be expected to be immune from various i
failures and malfunctions, regardless of the quality.of design, construction, j
and operation. Therefore 6 fu nher level of defense was established in that the plents were required to be designed to cos uccessfully with various equipment failures, transients, and postulated accidents. The scenarios for postulated accidents, to which all plants are designed to adequately respond, are known as design basis accidents and are detailed in the NRC's Standard Review Plan, which is used to evaluate the design of each j
nuclear power plant before the granting of a construction permit or an operating license.
Design basis accidents were chosen to represent a wide spectrum of f
plant problems, some of which were expected to be experienced in the plant's l
lifetime (such as failure of power systems), as well as events considered to be quite infrequent (such as major ruptures of piping systems) and not
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expected to occur in the plant's lifetime.
l The NRC Standatd Review Plan also identifies acceptable plant protection
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standards for each postulated plant accident. The requirements and capabilities of plant safety systems necessary to prevent these design basis accidents from leading to unacceptable radiological releases are specifically identified.
The Standard Review Plan give; acceptance criteria for judging the acceptability of the analytical results in response to these hypothetical scenarios. The resulting plant design incorporates smitiple and backup safety systems that will protect the reactor during a design basis accident and a postulated 3
L single failure in each system of these various protection devices.
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Notwithstanding the above, additional margins are required in the plant 1
design to protect the public even in the event of very unlikely accidents.
The reactor containment provides an additional level of safety. Design j
basis accidents for containment reflect a number of arbitrary accident l
t sequences developed from postulated events. For example, the containment structural design is based upon the effects of a concurrent earthquake and a rupture of major reactor coolant system piping. Concurrently, in order to j
assess the effectiveness of leaktightness, the safety systems are presumed not to be effective in cooling the reactor core, resulting in the. release of I -
fission products from the reactor core. Although the design basis accidents I
discussec above are allowed to result in some failed fuel (less than 1 percent), they do not result in significant core damage.
For the containment.
design, some independent failures of the protection systems are assumed to j
occur siriultaneously with the occurrence of the accident they are inhanded s
to control. Although the purpose of other safety systems is to shut down the reactor fission process and provide emergency cooling water to the reactor core, the containment has a required function of providing an l
. essentially leaktight barrier to " bottle up" any radioactive material released to the containment through any rupture or break in the reactor coolant system. Given the release of the radioactive material and cooling water, the containment is required to retain this material and prevent sirnificant releases to the environment. Consequently, the assessment of cortainment design adequacy assumes the postulated release of fission products to the containment irrespective of the performance of the core cooling safety systems.
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.c 10-Although design basis accidents are used to determine-the adequacy of plant systems' design and performance under postulated accident conditions.
severe accidents are analyzed by imposing a set of additional assumptions to further presume that these systems will not work as designed. The m
centainment design basis reflects a combination of parameters incorporating several design basis accidents for structural considerations coupled with an assumed release of radioactive material to containment for assessing leak-I tightness.
In summary, the design purpose of the reactor containment is to protect i
b against postulated radioactive releases from hypothetical reactor acci-dents up to and including major ruptures of reactor coolant piping, where such events resulted in some degree of core damage.
These hypothetical events postulated a release of fission products from the reactor core to the reactor coolant system and subsequently into the containment through the pipe break. This was considered one of the less likely, but possible acci-dents and supplied a straightforward means of providing additional margins for containment design.
The concept of severe nuclear accidents and how these accidents fit within the framework of protection from design basis accidents must also be considered.I For the last several years, the staff has been studying the likeliho,od and consequences of extremely low probability accidents involving multiple failures that lead to core damage.
This class of accidents is I Severe accidents are defined as those "in which substantial damage is done to the reactor core, whether or not there are serious offsite consequences."
This definition is extracted from the " Policy Statement on Severe Reactor Accie nts Regarding Future Designs and Exist' ; Plants," 50 Fed. Reg. 32138, August 8, 1985.
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, beyond the existing design basis and'is generally known.as severe accidents.
This evaluation was first done-comprehensively by the Reactor Safety Study (WASH 1400), which is known as a probabilistic risk assessment (PRA).
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types of accidents = s'udied in this evaluation are basically those in_.which backup safety systems fail, ev4ntually resulting in damage.to the nuclear
' fuel and considerable releases of radioactive material outside the. reactor.
.i p1 cooling system into the containment. Depending on other failures and p
p containment behavior, significant radiological releases into the environment q
could conceivably ' occur.
Implicit in these scenarios is.the development of 1
a better. understanding of containment performance and its failure mechanisms.
More detailed PRA studies have been conducted since the pubikeatien of l WASH-1400 to better under' stand the probability of these unlikely events and also-to better predict the magnitude of potential radiological releases into d
the environment, given a containment failure and attendant consequences Considerable work has also focused on the behavior of reactor containme~*.s 4
following a severe accident in which molten reactor fuel could potentie M melt through the reactor vessel.
Results of such studies have general'a confirmed the very low likelihood of such accidents and' the relative 13 :w risk to the public even if such very low probability accidents were to iI occur.. Although not originally designed to protect against some of the severe accidents, reactor containments provide consjderable protection due j
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to their ability to reduce radiological releases to the public from such accidents. For example, the results of research work indicate that the actual pressure-retaining capability of most containments is well above l
- their original design pressures.
Studies also indicate that the massive I
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material even if they were to fail following a core melt event. As' discussed h
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below, there exists a wide range of uncertainty regarding a hark I contain.
ment's behavior. during'a core melt accident. A recent study judged the
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ibility of. some form of containment failure, assuming a core melt.NN 4
occurred, to be between 10 and 90 percent.N However. the total core h
y dange freq'uency for the BWR Merk I desien (Peach Bottom) was less than the L
total core damage frequency of the other four reactor dirsigns. studied'by
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generally an order of magnituce or more, L
Because of the very complex procestas involved in a severe reactor accident, exac.t predictions of accident consequences are difficult. Consider-i L
able research is under way to provide additional information in this arte.
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Results from such studies allow NRC staff to focus attention on areas in which improvements can. be made to provide ' increased levels of safetylfrom O
these very unlikely events. The purpose of these projects is to conduct 1
hypothetical "what if" studies,= to understand ways public risk from neclear t1 i
, operations can be justifiably reduced. The results of our studies indicate' l
that risks from these severe accidents are very low and do not warrant immediate actions.
Petitioner has u.m essed concerns that are based on a memorandum-(
written on September 20, 1972, by Dr. S. H. Hanauer, a member of the staff oftheAtomicEnergyCommission(AEC)(theNRCsucceededtheAECin1975).
These concerns relate to the ability of the Mark I containment to respond The " Reactor Risk Reference Document" - Draf t (NUREG-1150), February 1987
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adequately to its original design function (i.e.,_ deal with a large loss-of 1
= coolent accident). Dr. Hanauer's memorandum raised seven concerns, all of which centered on the viability of the pressure-suppression containmenti
. concept. They relate to steam-bypass susceptibility, valve reliability, lack of: adequate testing, and volume limitations causing overcrowding, u
When Dr. Hanauer's seven conc 3rns were raised,Lthe staff evaluated each of them to determine whether adequate safety margins were being maintained-or. cxisting plan'ts.
Subsequently, the NRC staff concluded that Dr. Hanauer's concerns had been properly considered and documented its findings in ~
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NUREG-0474, "A Technical-Updete on Pressure Suppression Type Containments in
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Use in U.S. Light Water Reactor Nuclear Power' Plants," issued in July 1978.
Enclosure A to NUREG-0474 sumarizes NRC staff actions' related to each t
= of the seven concerns identified in Dr. Hanauer's memorandum of September 20, 1972. A copy of that enclosure is being provided to the Petitioner w'ith.
this Decision.
Each statement of concern was followed by a-response ~ that reflected the NRC evaluation.
In each case, the response showed that the 1
NRC no longer consicered the concern an unresolved safety issue.
It should be noted that although the concerns reflected the views of
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. Dr. Hant.uer in September 1972, the NRC response reflected the status of the issues in July 1978.- Moreover, by June 1978 Dr. Hanauer had changed his
. opinion regarding his 1972 concerns, as reflected in a memorandum dated June 20, 1978, in which he stated: "Thus while we may yearn for the greater simplicity'of ' dry' containments, the problems of bcth ' dry' and pressure-suppression containments are solvable, in my opinion, and the design safe, therefore licensable" (NUREG-0474).
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. 14-I Our review of the Petitioner's concern that is based on Dr. Hanauer's 3
-l memorandum indicates that this concern has been addressed in NUREG;0474.
Although=various changes have occurred since then, the fundamental safety J
conclusions stated in NUREG-0474 are essentially unchanged. The'most-not'able of the. changes has been the NRC position'related tu rendering the containment inert. E/ Since.NUREG-0474 was issued, the regulations relating to this issue (10 CFR 50.44, " Standards for Combustible Gas Control. System '
in Light-Water-Cooled Power Reactors") have been revised to require all Mark I and 11 containments to be rendered inert. The response to Dr. Hanauer's
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.j 10 CFR 50.44; the Comission required all Mark I and Il containments to be a
1 rendered inert to accomodate the degraded core accident. ' A eview of this L
an'd other changes made since NUREG-0474 was issued, indicates that in no case
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containedLin NUREG-0474.
Test programs were initiated by utilities owning Mark I plants as part L"
of a. program in response to NRC letters that were transmitted in February.
l and April 1975 to all utilities owning BWR facilities with Mark I design containments. The letters requested that the owners quantify the hydrodynamic p
and safety-relief valve (SRV) discharge loads and assess the effect of these l
loads on the containment.
(These loads had not been considered during, the E/ n inerted containment is one in which oxygen is replaced by enough nitrogen A
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- 4 15-i licensingofthe'individualplantsbecausethese. loads (includingpool, 71 swell)wereidentifiedintheperiod1972through'1974aspartcfthereview.-
> cf the large-scale testing of the Mark III containment system design.)
I As'a result of these letters from the NRC'and in-recognition that the:
r evaluation effort would be very similar for all Mark I BWR clants, the!
utilities formed an ad hoc Mark I Owners Group.
The objectives of this:
Owners Group were to determine the magnitude and significance of these
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dynamic loads as quickly as possible and to identify actions to resolve any.-
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09tstanding. safety concerns. A series of generic test programs wat' created p y r
to accomplish.these objectives.
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Since NUREG-0474 was issued in July 1978,~ the generic test programs
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I related to the Mark I containment design and the NRC assessment of the' tests l
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'have been completed. The staff evaluation of the generic test programs was j
Ll reported in NUREG-0661, " Mark 1 Containment Long Term Program Safety Evalua-a tion Report " issued in' July 1980..NUREG-0661 describes and presents' staff conclusions regarding the generic techniques for:the definition of suppression pool hydrodynamic loads in a Mark I system and the related structural acceptance criteria. As part of the acceptance criteria, the staff-required i
-that a_ plant-specific analysis be submitted 'by= the licensees for all 24 plants having Mark I containments. These analyses have.been reviewed and i
approved by the staff. All modifications proposed by the licensees to 1
satisfy the criteria contained in NUREG-0661 have been completed.
Another of Dr. Hanauer's concerns focused on the safety disadvantages y
a of pressure-suppression containments. This issue is related to the possi-i bility of steam bypassing the suppression pool in BWR pressure-suppression
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a containments, and wasJdesignated Generic Issue 61, "SRV Line Break Inside l
the Wet.Well Airspace of Mark'I and II Conta'inments.' 'An evaluation of. this q
~ issue ' as been completed, and'the results were presented in NURIG/CR-4594, h
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- Estimated Safety Significance of Generic Issue. 61," which,was issued in June 1986. On the basis of these results, the staff concluded that no new i
requirements were justified and no further study of.this safety' issue was -
warranted.
The Petitioner al' o raises concerns regarding the possibility that_ the
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BWR containments might' fail in the event of a severe accident. The Petitioner cites various studies regarding a high probability that Mark I. containment l
structures will not stand various severe accident scenarios.
4
'1 As discussed previously, the NRC views probabilistic risk assessment as-a structu ed method for investigating the likelihood and consequences of L
re6ctor accidents considered to have a very low frequency of occurre'nce.
The perceived inability of the Mark I containment.to survive a: severe accident has been postulated by the Petitioner as.a design flaw.
The evaluation of severe accident vulnerability involves three distinct evaluations. The first involve: the probability of an accident involving a
core damage, the second involves the likelihood of containment failure, and the third involves an assessment of the radiological consequences and public
. doses resulting from the accident. All three issues must be considered in making a determination on the magnitude of severe accident risk and the L
actions that should prudently be taken to reduce that risk.
The studies that have been conducted emphasize that their results 1
L inherently possess large uncertainties. The draft results of NUREG-1150
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properly-reflect the areas of uncertainty. This study included an evaluation for Peach Bottom, a plant quite similar in design to the typical Mark I reactor and containment. The study presented the estim6ted mean frequency of core damage ~'as approximately I chance in 100,000 per year of operation.
. Another comprehensive risk study conducted by the NRC staff estimated a mean -
4 core damage probability of 1 in 10,000 for the Limerick plant.
These results are consistent with NRC's belief that' core melt accidents
- are very unlikely.- ' Draft NUREG-1150 also investigated the' probaH11ty of ea'rly containment failure following a core melt and concluded that our ability to accurately predict the response of a Mark I' containment was limited for-situations in which it was subjected to the l'arsh temperature knd pressure conditions following a core melt accident. As stated earlier, the report indicated that containment failure probability (for these extremely unlikely events) could likely range from 10 to 90 percent.
-These uncertainties are currently the subject of research efforts to better predict the behavior of containment $ during severe accidents so that a more complete risk perspective can be assembled for guiding our regulatory activities. However, it is important that these uncertainties be properly characterized. They are not identified deficiencies in the BWR Mark I containments, which have been demonstrated to satisfy their design performance requirements. Rather, the:e uncertainties guide our research investigations, whose goals are to provide improved understanding of very unlikely risk situations at nuclear power facilities. Results from these studies (including l
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high containment failure probabilities) also allow us to calculate'public risk estimates assuming that one element of the three in a risk assessment i
7 (containment failure) is less favorable.
Even allowing the large uncertainties-that result in a high upper value
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for containment failure, the NUREG-1150 study estimated that the probability 1
of a large reactor accident resulting in one or more early fatalities ranged d
from 1 in'1 million to 1 in 1 billion.
In the event of a severe. accident.
both the probability of very high radiation exposures and the distances over
.c which such exposures would occur were estimated to be reasonably small. The-risk levels for each Mark I reactor would:of course depend on its actual-(
' core melt probability, containment behavior, the local demography, and could vary somewhat from the _ results presented in NUREG-1150. The results of this and related stud'ies do, however, support our overall conclusion of low severe accident risk at Mark I reactors. One contributing factor is that
= the massive reactor containment structure may retain considerable radioactive material following a core melt event even if its pressure boundary fails.
I
-t in this regard, containment failures include cracks or other phenomena that result in loss of pressure integrity that can result in le&ks but should not be viewed solely as catastrophic failure of the containment structure.
In4 the event radioactive material is released inside containment, some of this material dispersed in air, e.g. radiciodine, will be deposited on surfaces inside containment. Even though NRC analysis gives no credit for this phenomenon, deposition of material'within containments, even though there may be leakage, 'will increase the time available to implement effective protective action activities.
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19 Although we believe that severe accident risks are low at operating 3
' nuclear plants, to assure.that our risk conclusions are. applicable 'to all c.
L operating units, a number of programs are going forward'to assess severe
{
accident -likelihood and consequences. These-programs. include plant-specif1e studies to determine any severe accident vulnerabilities, both from the1 o
perspective of' accident frequencies and from containment performance following a core melt. Any problems will be dealt with if identifiec. One program is known as the Individual Plant Exam'ination (IPE) Program and is currently under way. This program and other related programs will be conducted to provide further assess wnts of severe accidents on a plant-specific basis so that appropriately low risk levels can be maintained.
Evaluations of the Mark I containment with respect to severe accidents
[
i arecontinuingthrough(1)theimplementationoftheCommissionPolicy Statement on Severe Accidents,- (2) the NRC staff and industry dialog'ue to improvecontainmentsevereaccidentperformanceforallBWRs,and(3)'the containment performance improvement program.- With respect to the latter program, the staff identified a number.of modifications that-substantially enhance the Mark I plants' capability to both prevent and mitigate the consequences of severe accidents. Theimprovementsidentifiedinclude(1)-
improved hardened wetwell vent capability (2) improved reactor pressure vessel depressurization system reliability, (3) an alternative water supply.
to the reactor vessel and drywell sprays, and (4) updated emergency procedures and training.
l After considering the staff's proposed Mark I Containment Performance Program the Commission directed the staff to pursue Mark I enhancements on e
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a plant-specific basis in order to account for vossible unique design
' differences that may bear on the necessity and nature of specific safety
- improvements. Accordingly, the Commission concluded that.the recommended-safety improvements, with one exception, hardened wetwell vent capability, J
should be evaluated by licensees as part of the Individual Plant Examina-tion Program. With regard to the recommended plant improvement dealing with' hardened vent capability, the Comission, in re' cognition of the circum-stances and benefits associated with this modification, has directed a
.different approach. Specifically, the Comission has directed the staff to approve installation of a hardened vent under the provisions of 10 CFR 50.59 1
for-licensees who, on their own initiative, elect to incorporate this plant improvement. The staff previously inspected the design of such a system that was installed by Boston Edison Comr.any at the Pilgrim Nuclear Power:
Station. The staff found the instal 16d system and the associated Boston Edison Company's analysis acceptable.
In response to the Comission's directive, the staff issued Generic Letter 89-16, " Installation of Hardened Wetwell Vent'," on September 1, 1989,-
to-all holders of operating licenses for nuclear power reactors with Mark I containments requesting I Mansees to submit their plans for addressing the hardened vent issue. Licensees were encouraged to install a hardened vent-under the provision of 10 CFR 50.59 or to provide installation cost estimate information in order that the staff may perform plant-specific backfit analyses.
As indicated in the discussion above on the Mark I containment, the Petitioner has not presented sufficient evid6nce to indicate that Mark I
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ik resctors should not operate whb e risk-reduc' tion improvements are being considered. That'is, there is not sufficient' evidence of either design flaws in Mark I reactors o* high risk to warrant suspending the operating licenses for those reactors.. Therefore, this portion of the Petitioner's s
request ~is denied.
B.
Mark 11. Containment. Concerns As stated above. Petitioner alleges that Mark 11 reactors, supposedly 3_
an improveme6t over the Mark I model, still have many possible scenarios for l
early containment failure according to NRC safety studies. Again, Petitioner
-i does not provide any information of which the staff was unaware. Much of what has been already stated in the discussion of the Petitioner's concerns I
with respect to Mark I containments as to containment-design, functional purpose, and performance during severe accident scenarios applies equally to Mark 11 containment types.
i The NRC'is currently studying Mark 11 containment performance. The study reviews challenges to the integrity of the BWR Mark Il containment thit I
could_arise from severe accidents. The challenges are organized into two' broad groups: those in which containment integrity is challenged before a
extensive' core damage, and those in which core melt occurs first, with containment integrity not threatened until the time of reactor vessel failure or later. Also reviewed are some proposed improvements that have the potential to either prevent core damage or containment failure,'or to mitigate the consequences of such failure by reducing the release of fission products, and thus the offsite consequences. For each of the proposed improvements, a preliminary qualitative analysis of the impact upon core melt frequency and risk has been performed.
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.22 Becauseofthelargephenomenologicaluncertaintiesandthestateof;
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flux'of the ongoing research efforts, the conclusions.about potential i
7' improvements are viewed as tentative. The estimated costs for selected improvements were taken from previously published information..They were' not meant to be interpreted as final estimates as no cost-benefit analysis was performed.
Among the potential improvements for the first category of containment ti challenges,are containment pressure control, such as venting from the'-
wetwell through a hardened vent pipe, and containment pressure control and fission product scrubbing,'such as the use of containment sprays with 'a.
- backup water supply, For the secondary category of containment challenges, proposed improve.-
s ments include containment pressure control, for example, a hardened vent from the wetwell;- improved means to depressurize the reactor, for example, 3
enhancements to the Automatic Depressurization System (ADS) and the safety.
t reliefvalves(SRVs);containmenttemperaturecontrolandfissionproduct
. scrubbirg, for example, containment sprays with a backup-water supply; j
enhanced operability of the suppression pool cleanup systems for removal of suppression pool water and enhanced operability of the reactor water cleanup-system for decay-heat removal and external cooling of the drywell head; and mitigation of the fission product release, for example, use of fire protection sprays to enhance fission product retention in the reactor building. As i
indicated previously in the discussion on Mark I containment performance, programs are also under way to evaluate Mark II containments for performance i
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a during; severe accidents'.- The.results of:these programs will:be evaluated in accordance with the Comission's regulations to detemine whether any improvements should be required as a backfit.-
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' As stated previously,' Petitioner has not presented sufficient evidence
- l to indicate that Mark II reactors should not operate while risk-reduction-
,1 improvements are being considered. That is, there is not sufficient evidence
'of.either design flaws at. Mark II reactors:or high risk to warrant suspending l
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the operating licenses for those reactors. Therefore, this portion of the-1 L
Petitioner's request is denied.
E C.
Additional-Reed Report Concerns j
The Petitioner also lists two concerns related to the 1975. General h
a Electric Company " Reed Report." These are, according to the' Petition, as 1
follows:
I L
I.
In 1975, General Electric engineers wrote an internal report highly d
critical of their own company's nuclear reactors. This Reed Report was.kept secret by both General Electric and the Nuclear Regulatory Commission unti1~
1987, when it was released under pressure by State and local governments in cooperation with safe energy organizations.
The General E1' ctric engineers
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detailed dozent of safety and economics problems with all the reactors, L
concluding t'.at General Electric reactors are "not a quality product." In l
fact, the engineers recommend that General Electric stop selling their Tl E
react?rs.
2.
The Mark II reactors, on which the 1975 General Electric Reed l
Report was primarily focused, have the aforementioned " dozens of safety and economic problems," and have suffered massive cost overruns during construc-
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tion as a result of design problems.
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i TheReedReportwasaself-criticalstudyperformedbythestaffofthe General. Electric Company in 1975.: It wasLintended'as aLproduct improvement
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study to enhance the availability and performance of GE's boiling water reactors. The report, by its nature a candid self-analysis, was intended for GE's internal use only.
It had always been held by 7 to be " proprietary" i
and thus was not subject to,public disclosure.
e The principal author of'the report was Dr. Charles E. Reed, a Senior:
Vice' President of GE. Contributors included technical and professional h
personnel from a variety.of GE departments. Their efforts resulted in the
}
Nuclear Reactor Study, referred to today as the Reed Report, and a set of 10L i
subtask reports that provided the detailed technical infonnation used to -
h develop the Nuclear Reactor Study. The Reed Report addressed operating BWRs and;the design of future GE products.and services in the nuclear field. -For rL3 tors in operation at the time, the report discussed ways to impr'ove a plant's availability and *ts electrical generating capacity factor through improvements in pir.t hardware and also in service, fuel,, equipment, and operating procer'.ures. For future reactors the report considered GE's r
- then-new BWR design, the BWR-3, and discussed problems regarding final' design details,-licensing, and full-power operation-of BWR-6 plants.
The NRC first learned of the existence of the Reed Report in a casual conversation between the NRC Chairman and one other Comissioner and GE a
officials at the San. Francisco airport on August 21, 1975. There was further mention of the report in the Congressional Joint Comittee on Atomic Energy hearings held in February and March 1976. At that time, Dr. Reed
. testified regarding the report.
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OnL February 23-24, 1976, two NRC staff members reviewed a copy of the 1
' report in GE's 1ashington, D.C., offices. They. determined 'that the report-
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(1)ldid not 1dentify any new safety concerns, and (2) did not indicate thatL
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-GE had failed to report any significant safety = concerns toLthe NRC.
On March 6.- 1978, in response to a request from Congressman John D.- Ding ill,.
h the NRC asked GE-to provide either a copy of the Reed Report or a list of the
=p safety issues it adoressed. On March 22, 1976, GE gave the NRC a~1ist of 25' issues idantified as having "some safety significance.": On May 26, 1978,'GE-
,provided to the NRC a safety evaluation of the 25 issues it had identified.
On November 9 1978, the NRC staff gave the Comission the' results of:
_a its updated review ofothe Reed Report ano found "no substantive disagreementa J
with,the summary status provided by GE."
The NRC first received a copy of the Reeo Report on Janusry 5,1979, 1
under a protective agreement, when GE gave a copy to the Atomic Safety and.
l Lieerising Board in the licensing proceedings for the Black Fox nuclear plant.
t GE continued to categorize;the report as " proprietary" and claimed that the
' document was' exempt from mandatory public disclosure.
The NRC then received several Freedom of Information Act (F0IA)-requests
-t for the Reed Report, beginning with a request dated September 26, 1979.- After reviewing arguments for and against granting an FOIA request and after con.
sultation with the Department of Justice, the Commission voted on October 9, 1980, to release the Reed Report to the public; however, on October 17, 1980, GE sued NRC, seeking to prohibit the release. On December 21, 1984, the U.S. Court of Appeals for the Seventh Circuit ordered a remand to the i
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t Comission for its decision whether to release the Report $/.
Subsequently, in July 1986, the Comission voted to withhold the Reed Report from public
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l disclosure. GE subsequently.. released the Reed: Report in-July 1987 -in a.
)
.two-volume' document titled "12 Years Later... An Update Report on the I
- Nuclear Reactor Safety Study." The updated report describes how earlier N
i NRC reviews of 1976 and 1978 confirmed how all safety issues mentioned in !
1 the Reed Report had been disclosed to the NRC _previously.
It also describes how the study was performed early in the BWR-6 (Mark III containment) design
' cycle and how the recomendations from that report were-implemented beforr 4
1-BWR-6 Mark III plants =went into operation.
L Nonetheless, as public interest in the " newly discovered". Reed Report U
1 heightened, and n6twithstanding their earlier reviews 'of the: document, on L
June 2,- 1987, NRC established a special t"k 7 to evaluate again the issues raised in the Reed Report, taking into account the increased knowledge about nuclear power based on engineering = studies and operational experience in the 12 years since the Reed Report was written.
i l
The purpose of this review was to place these issues in a 1987 perspec-tive to ensure that the NRC staff truly had been aware of all safety issues
~
-discussed within the report and that the issues were either resolved or programs were under way to address those issues not yet resolved.
t j
Thi., review produced three separate conclusions:
i (1) The Reed Report does not identify any. matters that would support a need to curtail the operation of any GE boiling water reactor plants now licensed.
Il Ceneral Electric Co. v. U.S. Nuclear Regulatory Comission. 750 F.2d 1394 (7th G1r.1984).
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.1T (2) The~ Reed Report does not identify any new safety issues of which'the staff was unaware.
I (3) Although certain issues addressed by the Reed Report are still beingL studied by the NRC and industry, there is'no basis for suspending:
plant operations while those issues are being resolved.
o Since_ knowledge of the' Reed Report became public in 1987, the staff has addressed nurarous Congressional.and private inquiries as to the impact of
-th^' issues raised in the report on public health and safety. -As stated e
~
previously, the Reed Report did not raise-any new issues of which the staff
.was unaware. Further, corrective actions.either had been implemented or were-N being implemented to resolve.those issues. The Petitioner has not presented any evidence or 'any new ~ issues identified by the Reed Report of 'which the
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- staff is unaware, re has the Petitioner presented any evidence calling into question the adequacy of the corrective actions _ implemented since th'e Reed Report was' issued. On this basis,-therefore, the Petitioner?s request is denied.
D.
Economic Issues o
c, Insofar as Petitioner asks for relief because of " economic problems" or
'" massive cost overruns during construction as a result of design problems,"
l the NRC is without, jurisdiction to grant relief. The NRC has authority to a
y-T govern any activity authorized pursuant ~ to the Atomic Energy Act of 1954,
.as amended, in order to protect health and to minimize danger to life or property. Because economic problems and cost overruns raise no threat to public health and safety, they do not provide the NRC with a basis on which to act. Accordingly, insofar as Petitioner bases her request on economic or
' cost considerations, the Petition is denied.
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IV.- CONCLUSION-The Petitioner seeks the institution of a show cause proceeding pursuant g
to 10 CFR 2.202 to modify or revoke the operating license' of all BWR facilities.
s
- Failing that, the Petitioner seeks,' without specificity, to "fix' all hWR facil-:
ities.
i The institution of proceedings pursuant.to'30 CFR 2.202 is appropriate only where substantial health and safety issues-have been raised.. See x
Consolidated Edison Company.cf-New.Ycrk (Indian Point. Units 1, 2, and 3),
CLI-75-6, 2 NRC 173 (1975) andWashipoton.PublicPower. Supply.Sygm(WPPSS
^
NuclearProjectNo.2),DD-84-7,l19NRC899,923(1984). This.is the standard-that F have~ applied to the concerns raised by the Petitioner in this decision' i
to.catermine whether enforcement action is warranted.
1 For the reasons discussec'above, I conclude that nu substantial health and $6fety issues h6ve been raised by the Petitioner. Accordingly,k,he z
Petitioner's request for action pursuant to 10 CFR 2.206 is denied.
I
' As' provided in 10 CFR 2.206(c), a copy of this Decision will be filed with the Secretary of the Commission for the Commission's review. The Decision will:
0 become final action of the Connissian twenty-five (25) days after issuance unless the Commission on its own motion institutes review of the Decision within that time, m
j J
FOR THE NUCLEAR REGULATORY COMMISSION 7
1 Thomas E. Murley, Director Office of Nuclear Reactor Regulation
-' Dated at Rockville, Maryland, this 4th day of December 1989.
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q 7590-01 UNITED STATES NUCLEAR REGULATORY COMMIS$!0N l
i" DOCKET W0. 50 293, et al.*
BOSTON EDIS0N COMPANY, et 41.*
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-(Pilgrim Nuclear Power Station et al.)*
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ISSUANCE OF DIRECTOR'S DECISION UNDER 10 CFR 2.206
-(;.
. Notice is hereby given that the Director, Office of Nuclear Reactor
'Regulatie'n-(NRR),hasissuedaDirector'sDecisionconcerningaPetition>
d
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dated. March 8,1989, flied by Ms. Anna Harlowe. Issues Coordinator, on, behalf.of the Ecology Center of Southern-California. The Petition asked f
7 the Director, NRR,.to take action to relieve what the Petitioner alleged p
to be undue' risks to the public health and safety posed by the containment I
design of boiling water reactors (BWRs), as revealed by various NRC staff members' statements, published studies, and by the 1975 General Electric
" Reed Report." The specific relief requested was to order all'BWR licensees D'
to,"f'ix"'or close all BWR reactors. Ms. Harlowe gave as grounds for the
. Petition that (1) in 1972, a member of the NRC staff recommended that GE J
p c%signedreactorsbebannedin-theUnitedStates;(2)-in1975,GEengineers.
a genrated the " Reed Report" that detailed dozens of safety and economic-
-problems with.GE-designed reactors and recommended that GE stop selling _
those. reactors;-(3)in1986,anNRCofficialadmittedthat24GEreactors
'+
1 with Mark I containments had a.90 percent chance of failure in a nuclear accident;(4)in1987,an'NRCtaskforceconfirmedthatMarkIcontainments were virtually certain to fail in an accident;-(5) according to NRC safety studies, Mark II reactors have many.possible scenarios for early containment i
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. failures; and-(6) Mark 11 designs, on which the Reed Report focused, have dozens of safety and economic proolems and have suffered massive cost overruns during construction as a result of design problems.
,On June 5, 1989, the Director, NRR, acknowledged receipt of the Petition.
He informed Ms. Harlowe that (1) the Petition would be treated under 10 CFR 2.206 of the Comission's regulations, and (2) appropriate-action would be taken within a reasonable time.
The Director has now determined that Ms. Harlowe's requests should be,
denied for the reescns bat forth in the " Director's Decision Pursuant to 10
~ CFR 2.206"- (DD 9 ).
The Decision is available for inspection and
. copying in the Comission's Public Document Room, Gelman Building, 2120 L Street, N.W., Washington, D.C. 20555, and at the Local Public Document Rooms near the facilities listed below. The andresses and hours of: operations for the local public document rooms may be obtained by calling the following
/ toll-free number:
1-800-638-8081.
' A copy of the Decision has been filed with the Secretary of the Commission for the Comission's review in accordance with 10 CFR 2,206(c). As prov'ided-in 10 CFR 2.206(c), the Decision will become the final action of the Comission-twenty-five (25) days after issuance unless the Commission on its own motion institutes review of the Decision within that time.
FOR THE NUCLEAR REGULATORY COMMISSION
[
Thomas E. Murley, Director Office.cf Nuclear Reactor Regulation Dated at Rockville, Maryland, this 4th of December 1989.
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' CA',0 LINA POWER'& LIGHT. C0. (Brunswick Steam Electric Plant,~ Units.1 and 2,
=Dscket Nos. 50-324 and 50 325)-
' LEVELAND ELECTRIC'!LLUMINATING C0., ET _AL., (Perry Nuclear Power Plant,.
- n Unit'1. Docket'No. 50-440).
-COMMONWEALTH EDIS0N C0. (Dresden Nuclear Power Station,= Units 2 and.3, Docket Nos. 50-237 and 50-249)..(Quad Cities Station Units 1 and 2f Docket =
m Nos.' 50-254 Land 50-265), (LaSalle County Station,-Units l'and 2, Docket Nos.-
)
CONSUMERS POWER CO.- (Big Rock Point Nuclear Plant, Docket No. 50-155).
DETROIT. EDISON CO (Enrico Fermi Atomic: Power Plant, Unit 2, Docket No..
~!
50-341):
GENERAL PUBLIC UTILITIES (0yster Creek Nuclear Power Plant, Docket No. 50-219)
GEORGIA POWER CO. (Edwin I. Hatch Nuclear Plant, Units 1 and.2. Docket Nos.1 i
50-321.and 50-366)
GULF STATES UTILITIES CO. (River Bend Station,' Docket No. 50-458):
ILLINDIS POWER CO.- (Clinton Power Station, Docket No.= 50-461)
IOWA ELECTRIC LIGHT & POWER CO. (Duane Arnold Energy Center, Docket No.' 50-331)-
?
LONG ISLAND LIGHTING CO. (Shoreham Nuclear Power. Station, Docket No. 50-322)-
4
. MISSISSIPPI POWER & LIGHT CO. (Grand Gulf Nuclear Station Docket No. 50-416)-
4 L
NEBRASKA PUBLIC. POWER DISTRICT (Cooper Nuclear Station, Docket No. 50-298) '
l
-NIAGARA MOHAWK POWER CORP. (Nine Mile' Point Nuclear Station Units 1 and 2,.
i Docket Nos. 50-220 and 50-410) i NORTHEAST UTILITIES (Millstone Nuclear Power Station, Docket No. 50-245).
NORTHERNSTATESPOWERCO.(MonticelloNuclearGeneratingPlant,DocketNo.-
t 50-263 PENNSYLVANIA POWER & LIGHT CO. (Susquehanna Ste'am' Electric Station, Units 1 and 2, Docket Nos. 50-387 and 50-388)
PHILADELPHIA ELECTRIC CO. (Peach Bottom Atomic Power Station, Units 2 and 3, Docket Nos. 50-277 and'50-278). (Limerick Generating Station, Unit 1, Docket 0
No. 50-352)
POWER AUTHORITY OF THE STATE OF NEW YORK.(James A. Fitzpatrick Nuclear Power. Plant, Docket No. 50-333)
PUBLIC SERVICE ELECTRIC & GAS CO. (Hope Creek Nuclear Station, Docket No.
50-354) i TERNESSEE VALLEY AUTHORITY (Browns Ferry Nuclear Power Station, Unilis 1, 2, L
and3,DocketNos.'50-259,50-260,and30-296) l-VERMONT YANKEE NUCLEAR POWER CORP. (Vermont Yankee Nuclear Power Station, L,
DocketNo.50-271)
WASHINGTON PUBLIC POWER SUPPLY SYSTEM (WNP Unit 2, Docket No. 50-397) i b
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.8 NUREG-1283 1
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NRC Staff Evaluation of the
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General Electric Company 1 (Nuclear Reactor Study 7
" Reed Report")
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i NDTICE t
Aveliability of Reference Meterials Cited in NRC Publications Most documents ched in NRC publications will be eveliable from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.W.I Washington, DC 20665 =
3.~ The Superintendent of Documents, U.S. Government Printing Office, Poet Office Sox 37082, i
Washington, DC 20013 7082
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- 3. The National Technical information Service, Springfield, VA 22181 l-L Although the listing that follows represents the majority of documents cited in NRC publications, h is not intended to be exhaustive.
p Referenced documents available for inspection and copying for a fee from the NRC Public Docu-I ment Room include NRC oorrespondence and internal NRC memorands: NRC Office of Inspection and Enforcement bulletins, circulars, information notices,' inspection and investigation notions:
License Event Reporu: vendor reports and correspondence: Commission papers;and applicant and :
I licensee documents and correspondence.
The following documents in the NUREG series are available for purchase from the GPO Sales Propam: formal NRC staff and contractor reports, NRCsponsored conference proceedings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of
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, facteral Mapulations, and Nuckar Repuletory Commission lasuenose.
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- Document ~ evailable; from the National Technical-information Service.loclude NUREG series' f
reports and technical reports prepared by other federal agencies and reports prepared by the Atomic -
Energy Commision, forerunner agency to the Nuclear Regulatory Commission.
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' Documents evallable from public and special technical libraries include all open literature items, i
such as books, joumal and periodical articles, and transactions. Feders/ Megister notices, federal and
' state legislation, and congremional reports can usually be obtained from these libraries.
L Documents such as theses, dissertations, foreign reports and translations, and non NRC conference
' proceedings are available for purchase from the organization sponsoring the publication cited.
i Singte copies of NPC draft re;, orts are available free, to the extent of supply, upon written request to the Division of Information Support Services, Distribution Section,' U.S. Nuclear Regulatory
? Csmmission, Washington DC 20555.
Copies of industry codes and standards used in a substantive manner in the NRC regtlatory process L
are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are avallable li there for reference use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the o
American National Standards Institute,1430 Broadmy, New York, NY 10018.
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p NRC Staff Evaluation of the General Electric Company J J(Nuclear Reactor Study
" Reed Report")
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Manuscript Completed: July 1987 -
Date Published: July 1987 Office of Nuclear Reactor'Regulailon' I
U.S. Nuclear Regulatory Commission Washington, DC 20555 N
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ABSTRACT
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In 1975, the General Electric Company (GE) published a Nuclear Reactor Study, i
also referred to as "the Reed Report," an interna 1' product-improvement study.-
.i GE considered;the document " proprietary" and thus,.under the regu13tions of
.the Nuclear Regulatory Commission'(NRC), exempt from mandatory public disclo-sure.. Nonetheless,' members cf the NRC staff reviewed the document in 1976 and s-determined that it did not raise.any significant new safety ~ issues. The staff also reached the same conclusion in subsequent reviews..
However, in' response to recent inquiries about the report.:the staff.re-d evaluated the Reed Report from a 1987 perspective.
This re-evaluatioh, docu-r mented in this staff report, concluded that (1) there are no issues raised in e
the Reed Report that support a need to curtail the operation of any GE boiling water reactor (BWR);.(2) there are no new safety issues raised in the Reed Report of which the staff was unaware; and-(3).although certain issues' addressed by the Reed Report are still,being studied by the-NRC and the industry, there is no basistfor suspending licensing and operation of GE BWR plants while these-issues are being resolved.
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.I NUREG-1285 iii
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-TABLE.0F CONTENTS mx Pg ABSTRACT................................................................
iiiL T AB LE - 0 F CONTE N15......................................................
" y EXECUTIVE
SUMMARY
D......................................................
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' INTRODUCTION......................................................
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BACKGROUND........................................................
. 5' J2.1' History of the. Reed Report...................................
5:
2.2' Structure and Contents of the Reed Report....................
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2.3 ' History of NRC Actions Regarding the Reed Report.............
64 2.4 NRC Catacorization of' Reed Report Issues...................... 2.5 Recent NkC Actions Regarding the Reed Report'.................
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'3 THE " TWENTY-FIVE LICENSING ISSUES" IDENTIFIED BY GENERAL ELECTRIC..........................................................
' 9' 3.1 Degree of Completion of BWR-6 Design......................... :
9-3.2 Amount.cf. Margin Between Design Calculations for Core and Operating Limits.............................................
10.
3.3 Impact cf. Cold Shutdown Reactivit
. De s i gn...........................y Margi n on BWR-6 Core "E
10 4............
'3.4 Impact of'End of Cycle (EOC) Scram Rehetivity Insertion Rate on Core F ul l' Power. Li f e..................................
11
- 3. 5 - Long-Term Effect of Radiation in, Core Internals..............
11 3.6 Degree ~of: Proof of Accuracy of Transient Design-Methods......
13 s'
3.7 Impact:on Fuel Integrity of Reduced Moderator Tem due toLEquipment Failure........................perature 13' 3.8* Performance of-Relief' Valve Augmented Bypass (REVAB) System'..
14 3.9 Impact of: Hydrodynamic Phenomena on Containment Designs'......
14 3.10 Radiation Exposure from Removal-of Steam Dryer / Separator:
Assembly.....................................................
15 3.11 Level of Testing of Mark III Cqntai nment.....................
16' 3.12-Presence of Detectable Plutonium-Inside the BWR Turbine......
16, 3.13'The Effect of $1oshing of the-Suppression Pool on Mark III Steel Containment' Structure Design...........................
17 3.14 Evaluation of Fuel Transfer Accident in Mark III
" LN Containment...................................................
. 17' 3.15' Impact of Core Design and Licensing Criteria on BWR Capacity.....................................................
19.
3.16 Adequacy of' Design Procedures To Ensure Compliance with Licensing Criteria........................;.................
19 3.1' Consistency of Degree of Verification of Calculational Models........................................................
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-3.18 Possibility of Control Rod Binding Due to Fuel Channel Creep........................................................
20 3.19 Compliance of Design Work and Reviews with Written Procedures...................................................
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TABLE OF CONTENTS (Continued) o 3-
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' "i 3.20 Absence of Availability Goals in Design Procedur 3.21 Seismic Capabilities-of 8 x 8 Fueix $ pacer.'..... es...........
21-13.22 Extent of. Life of Position Sensor in Traversing In-Core -
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' o Probe ~ System................................................
3.23 Radiation Levels outside Biological Shield-and Drywell......
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st 3.25 Peak Pressurt y 'n efWS Calculations-for BWR 23 c
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'OTHER SAFETY-Siel,qr!CAA
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.THE RE2D REPORT...........SES~ IDENTIFIED BY THE NRC STA
...................................... - 25i, 4.1:. Combination cf LCik Induced Loads and Safety' Relief Valve.
'I 4.2 ; Jet Impingement.on.the Weir / Pool in a BWR Mark I 125 Containment..................................................,-
26-rx HS 4.3 Main Steam Isolation Valve Leak: Tightness....................
.t 14.4 Control of Design of Purchase Components.............
26
- 4. 5 Flow-Induced Vibration of Jet Pumps!.......................... :28 i
4.6 Stress Corrosion Cracking in Stainless Steel Piping........
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'29-5 THE GENERAL ELECTRIC SUBTASK GROUP REPORTS...
uo 13 1 5.1 Subtask-A:' Report on Nuclear S
- 5. 2: Subtas k B:
Report on Fuel...ystems.........................
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Report on Electrical C
and Instrumentation Systems.........,..ontrol:
5.4 Subtask D:. Report on Mechanical Systems and Equipment.......
36 5.5 Subtask E: -
42 Reportaan Materials Processes and Chemistry....
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Subtask Fi Re Construction.portfon Production,, Procureme,nt,.and.
47 5.7 Subtask.G:
Report on Quality Control System Overview.........
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- 5. 8 Subtask H:
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Report on Management /Information Systems H
15.9. Subtask I:.
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Report on Regulatory Considerations.............
5.10 Subtask J:
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Report on Scope.and Standardization.............
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F EXECUTIVE SUM ARY' o
The purpose of this NRC st'aff evaluation of the General Electric Nuclear,Reac tor Study (the Reed Report) and its 10 subtask reports.is to reconsider'the
- issues and concerns identified in-the report in the light of current knowledge, R
recent operating experience, and regulatory issues as they have developed since the report was issued in 1975; F,
'A Histo *vjf the Reed Report, j
The Reed Report was-a self-critical study performed by the~ staff of the General Electric Company (GE) in 1975. :It' was. intended as a product-improvement. study to enhance the availability.and performance of GE's boilin
- (BWRs). - The, report, by its nature a candid self-analysis,g water reactors was' intended'for GE's s
internal use only.
It has always been, held by GE to be " proprietary," and thus not-subject to public_ disclosure.
The principal author of the report was-Dr. Charles E. Reed, a Senior Vice Presi-
~ dent of GE.
Contributors included' technical and professional personnel-from a l:
variety of_GE departments.
Two products resulted from their efforts.- One was 3
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- the Nuclear.Reartor Study, referred to today as the Reed Report;_the second was.>
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aLset of-10 subtask reports that provided the detailed technical information 4
L used to-develop the Nuclear Reactor Study.
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The Structure of the Reed Report
'l The Reed' Report addressed operating BWRs and the design of future GE products
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.and services in the nuclear field.
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- For reactors in operation at the time, the report discussed ways to improve plant-W availability and its electrical generating capacity factor through improvements-in plant' hardware and through improvements,in service, fuel equipment, and i
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operating procedures.
Forfuturereactors,thereportconsIderedGE' sthen-new-BWR design, the BWR-6, and discussed problems regarding final-design details, L
111 censing, and full power operation of BWR-6 plants.
The report addresse'd 10 general topics, as follows-4 e
(1) nuclear systems (2) fuel (3) electrical, control, and instrtmentation (4) mechanical systems and equipment L
(5) materials, processes and chemistry
-(6) production, procureme,nt, and construction l
b (7) quality control systems overview (8)' management /information systems (9) regulatory considerations L
(10) scope and standardization NUREG-1285 1
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V Each of these general topics was addressed in a separate subtask report, and w
the 10 subtask reports were used to generate the Reed Report.
Nistory of NRC Actions Recardino the Reed ReDort The Nuclear Regulatory Commission (NRC) first learned of the existence of the Reed Report in a casual conversation between the NRC Chairman and one other Commissioner and GE officials at the San Francisco airport on August 21 There was further mention of the report in the Congressional Joint Committee 1975.
on Atomic Energy hearings held in February and March 1976.
At that time, Dr. Reed testified regarding the report.
On February 23 24, 1976, two NRC staff members reviewed a copy of the report in GE's Washington DC offices.
identify any new sa,fety. concerns and (2) did not indicate that GE h to report any significant aafety concerns to the NRC, On March 6, 1978, in res p nse to a request from Congressman John D. Dingell, i
the NRC asked _GE to provide either a copy of the Reed Report or a list of the-
'i safety issues it addressed. On March 22, 1978 GE gave the NRC a list of 25 issues identified as having "some safety significance."
i On May 26, 1978 GE provided to the NRC a safety evaluation of the 25 issues it had identified.
On November 9,1978, the NRC staff gave the Commission the results of its up-t dated review of the Reed Report and concluded:
with the summary status provided by GE."
"no substantive disagreement L
L The NRC first received a copy of the Reed Report on January 5,1979, under a Board in the Black Fox proceedings. protective agreement, when GE g "proprietar GE continued to categorize the report as disclosure.y" and claimed that the document was exempt from mandatory publit i
The NRC then received several Freedom of Informati6n Act (FOIA) requests for the Reed Report, beginning with a request dated September 26, 1979. After reviewing arguments for and against y
the Department of Justice, granting a FOIA request and after consultation with the Commission voted on October 9, 1980, to release s
the Reed Report to the public; however, on October 17, 1980, GE sued NRC, seek-ing to prohibit the release.
On December 21, 1984, for the Seventh Circuit ordered a remand'of the Commission's decision.the U.S. Cou quently. in July 1986, the Commission voted to continue to withhold the Reed Subse-Report from public disclosure.
Reed Report to the public.
To date, the Commission has not released the NRC Cattoorization of Reed Report issues On the basis of its reviews of the Reed Report and on information on the report supplied by GE, in November 1978 the staff grouped the 25 issues addressed in the report into six categories as follows:
constraints on operation resulting from regulatory requirements (7 items) plant-specific matters to be resolved in plant-specific license reviews (4 items)
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features already deleted from GE design (1 item) quality assurance issues (2 items)
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issues for which final resolution was pending, but for which interim posi-
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tions provided an adequate basis for allowing continued licensing of plants (8 items) issues already resolved by staff review (3 items)
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Recent NRC Actions Renardino the Reed Reoort On June 2' 1987, NRC established a special task group to evaluate again the issuesraIsedintheReedReport,takingintoaccounttheincreasedknowledge about nuclear power based on engineering studies and operational experience in the 12 years since the Reed Report was written.
This review produced three separate conclusions:
(1) The Reed Repbrt does not identify any matters that would support a need to
. curtail the operation of any GE boiling water reactor plants now licensed.
(2) The Reed Report does nut identify any new safety issues of which the staff was unaware.
(3) While certain issues addressed by the Reed Report are still being studied by the NRC and industry, there is a basis for permitting continued plant operations while those issues are being resolved.
e NUREG-1285 3
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kM 1 I 1WTRODUCTION
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(; d' The purpose of this WRC staff evaluation of the General Electric Company's l-Nuclear Reactor fet.udy (the Reed Report) is to reconsider the issues and con-I cerns identified in the report in the light of current knowledge, more recent
. plant operating experience, and regulatory issues as they have developed since
.the report was issued in 1975.
This re-evaluation was prompted by concerns expressed by public officials and others regarding alleged serious weaknesses in the safety of General Electric (GE) botiing water reactors (SWRs).
These statements of concern were reactions to recent accounts in the news media, particularly newspaper accounts, of a
" secret" GE report written in 1975.
The report referred to in news accounts is
.the GE Nuclear Reactor Study, which is more commonly called the Reed Report because Dr. Charles E. Reed, a Senior Vice President of GE, headed the task group whose studies culminated in the issuance of the Nucle' deactor $tudy.
Because of the nature of the study, GE has always held th6 need Report to be proprietary, not to be disclosed to the public or to GE's competitors, :The NRC has a copy of this GE proprietary report, along with the proprietary subtask reports and related material.
In the course of performing its regulatory func-tions, the NRC receivas and holds for review and for reference many proprietary documents from GE and.from other vendors of nuclear related products.
The NRC-staff.had long been aware of the Reed Report and its contents.
. Recently, however, in the discovery process of a lawsuit involving GE and the owners of the Zimmer facility excerpts from the Reed Report nal GE documents, apparently w,ere included in documents being, and other inter-exchanged between the parties in the lawsuit.
This material came into the possession of a news-paper, which purportedly disclosed some of the contents in a news article.
Some newspaper articles contained tecounts that stated or implied that the NRC had conspired with GE to keep this " secret" report from the public because of information that would be damaging to GE if it were disclosed.
These articles, together with interest from Congress, officials froiu the State of Ohio, and concerned citizens view re evaltation,of the Reed Report and the 10 subtask reports. prompted th The results of this current NRC staff evaluation are the subject of this report.
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J 2 SACKGROUND.
i 2.1 History of tht Reed Report i
i The Reed Report was a self-critical study performed by the staff of GE in 1975, with the stated objectives of " determining the basic requirements for im-pienenting the Nucleer Energy Division's (NED) quality strategy through con-tinuing improvements in the availability and capability of Boiling Water Reac-l ter Nuclear Plants (BWRs)."
l The principal author of the' report was Dr. Reed.
Contributors included tech-nical and professional personnel from a variety of GE departments.
Two pro-l ducts resulted from their efforts.
One was the Nuclear. Reactor Study, referred to as the " Reed Report"; the second was a set of 10 subtask reports that pro-i vided the detailed technical information used to develop the Nuclear Reactor i
Study.
The Reed Report was intended to be an internal document, not one for public dis-closure because, as claimed by GE, it contained information and comments that
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could have an adverse effect on GE's market position with respect to its competitors.
l Although GE allowed NRC to review the document on several occasions and eventually.provided NRC with a copy, GE also sued NRC to prevent the agency l
- from releasing the document to the public.
L 2.2 Structure and Contents of the Reed Report I
The report addressed 10 general topics related to the GE nuclear power product 4
line; these topics were:
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(1) nuclear systems (2) fuel (3) electrical, control,'nd instrumentation a
(4) mechanical systems and equipment i
,(5) materials, processes, and chemistry L
(6) production, procurement, and construction (7) = quality control systems overview (8) managament/information systems (9) regulatory considerations (10) scope and standardization
-Each of these general topics was addressed in a separate subtask report, and the 10 subtask reports were used to generate the Reed Report.
The subtask reports are discussed in detail in Section 5 of this report, t
For reactors in operation at the time, the report discussed ways to improve plant availability and its electrical generating capacity factor through in-provements in plant hardware and through improvements in service, fuel, equip-ment, and operating procedures.
For future reactors, the report considered GE's then-new BWR-6, and discussed problems regarding final design details, licensing, and unrestricted full-power operation, i
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J 2.3 History of NRC Actions Renardina the Reed Reoort 1975-1976
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The NRC first learned of the existence of the Reed San Francisco airport on August.21, 1975.
According to testimony given at the i
Hearing of the Joint Committee on Atomic Energy in its Investigation of Relating to Nuclear Reactor Safety February 18, 23 1976, the mention of the report was oral and very ge and 24 and March 2 and 4, i
neral in nature.
However because concerns we?e raised about the contents of the Reed Report, February, 23-24, 1976 'two NRC staff members reviewed a copy of the report in GE's !
Washington, DC office,s.
any new safety concerns of which the NRC was'not aware, andTh i4 the requirements of Section 206 of the Energy Reorganization (Act of 1974 in2) if!
regard to the reporting of significant safety items.
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On the basis of their review, these staff members did not identify any new s
safety concerns or an reported to the NRC y evidence that significant safety concerns had not been i
A copy of their memorandum to the Director cf the NRC Office of Nuclear Reactor Regulation (NRR) that documented their conclusions l
was incorporated into the record of the hearing of the Joint Committee on
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Atomic Energy's investigation of Charges Relating to Nuclear Reactor Safi 1977-1978 l
In December 1977, Congressman' John D. Dingell asked the Commission to
.l information on the Reed Report.
The Chairman responded in a letter dated 1
February 9,1978, which described the staff's earlier review and its conclusions i
To provide further information to the Congressman, on March 6, 1978, the NRC asked GE to provide either a copy of the Reed Report or a list of the safety issues.it addressed. GE responded by a letter dated March 22, 1978, which t
I contained a list of 25 issues identified as having "some safety significance."
On April 11, 1978, Dinge11's staff reviewed the report itself at the GE offices in DC.
And, on May 26, 1978, on each of the 25 items.
GE sent a letter to NRC that gave a status report On November 9,1978, the NRC staff gave the Commission the results of its dated review of the Reed Report (SECY-78-462A.
"no substantive disagreement with the summary) status provided by GE."T staff also grouped the 25 issues in the report into six categories.
The In a letter dated December 27 1978 and conclusions to Congressman, Dinge,ll.the Chairman forwarded the staff's findings 1978-1979 On October 18, 1978 the Atomic Safety and Licensing Board (ASLB) in the Black Fox proceedings issu,ed a subpoena to GE calling for GE to provide a copy o NUREG-1285 6
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' Reed Report for the proceedings.
GF. refused, claiming the report was " pro-prietary " and thus protected from mandatory public disclosure under the Commission's regulations.
GE and the ASLB were able to settle on the terms of a protective agreement, under which GE provided a copy of the report on January 5,1979.
This was the first time NRC had a copy of the report.
However, under the terns of the protective agreement, the report itself was never introduced into the Blaci. Fox proceedings The protective agreement did allow the following:
i (1) The Reed Report was made available to the ASLB in confidence.
(2) Verbatim extractions from the report were available to counsel insofar as they related to the Intervenor's contentions and the ASLB's questions.
(3) The report was available to the Intervenor's counsel to evaluate the faithfulness of the extractions.
The parties also signed protective agreements that limited access to and use of the report.
In September 1979, the NRC received the first of several F0IA requests for the Reed Report.
1980-1984 Several FOIA requests for the Reed Report were received in this period, the first actually having been made in September 1979.
On October 9, 1980, after hearing arguments on a request made under the FOIA, the Commission voted to release the Reed Report to the public.
However, on October 17, 1980 GE sued NRC, seeking to prohibit the release.
Subsequently, on December 21, 1984 the U.S. Court of AppealsfortheSeventhCircuitordereda.remandoft,heCommIssion's. decision.
1986-1987 In July 15E6, the Commission voted to continue to withhold the Reed Report from public disclosure.
This decision was based on the Commission's desire to en-courage similar studies and ensure NRC access to their results.
On June 3, 1987, Ohio Citizens for Responsible Energy (OCRE) filed suit in Ohio Federal District Court seeking public release of the report under the Freedom of Information Act.
To date the Commission has not released the Reed Report to the public.
2.4 NRC Categorization of Reed Report Issues In its November 1978 report to the Commission (see above), the staff grouped the 25 issues addressed in the Reed Report into six categories as follows:
constraints on operation resulting from regulatory requirements (7 items)
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- specific matters to be resolved in plant specific license reviews (4 items) features already deleted from GE design (1 ites) s
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quality assurance issues (2 items) e issues for which final resolution was pending,-but for which interim posi-i (8 items)tions provided an adequate basis'for allowing continued licensing 4
s q
issues already resolved by staff review (3 items) 2.5 Recent NRC Actions Regardino the Reed ReDort On June 2, 1987, following the appearance of newspaper stories with contro-versial accounts of the contents and safety implications of the. report, and statements attributed to some public officials and others in these newspaper 6
accounts and the receipt of inquiri*es from Congress, NRC established a special I
task group to re-evaluate the issues raised in the Reed Report account the increased knowledge and understanding of. nuclear po,wer issuestaking into I
gained in the 12 years since the Reed Report was written.
Martin Virgilio was appointed task group leader. Other people were named as needs were identified-for specific expertise.
effort are listed below. The people who contributed significantly to this Martin Virgilio - task group leader 1
L Rcby Bevan - technical coordinator Ed Shomaker - legal: counsel C. Y. Cheng - technical expert Tim Colburn project manager John Craig - technical manager, Perry Nuclear Power Plant l i
. Walt Haass - technical expert 1
Warren Hazelton - technical expert Wayne Hodges - technical manager Jack Kudrick - technical expert Oliver Lynch - technical expert Jerry Mauck - technical expert r
Robert Pettis - technir.a1 expert 1
h Laurence Philli John Ridgely ps - technical expert technical expert Chen Tan - technical expert John Thoma - technical expert Charles Tinkler - technical expert Robert Wright - technical expert e
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3 THE " TWENTY-FIVE LICENSING !$$UES" IDENTIFIED BY GENERAL FLECTRIC As discussed above, the Reed Report was not concerned primarily with safety-issues associated with GE BWRs, but with plant availability and electric gen-erating capability and, hence, the marketability of the GE nuclear reactors.
I However, in response to requests by NRC, in 1978. GE's Nuclear Safety and Licensing organization reviewed the report and identified 27 safety related items.
The 27 issues were subsequently consolidated into 25 when 2 of the items identified earlier were included under other issues.
The NRC staff has again reviewed these 25 issues in the light of current knowl-edge of nuclear safety, and the results of that review are given below.
For each of these issues, there is a statement of the issue, a statement of its safety significance, and a statement of the current status of.the issue.
The staff finds that none of the 25 issues identified by GE as having some safety significance involve any safety considerations not already identified and appropriately addressed by the staff.
3.1 Decree of Completion of BWR 6 Desian Issues The Reed Report noted the following with regard to the BWR-6 Mark !!! design:
i (1) The BWR 6 Mark !!! design was ine 91ete (in 1975), and several important technical problems were unresolvs;.
(2) The overall design of the BWR-6 Mark !!! is not well integrated.
The design was a result of a process of evolution and reaction to competitive offerings and regulatory requirements.
i (3) Future potential problems in the areas of fuels management, operational limitations, licensing, and, component replacement had to be anticipated.
Safety Sionificance None.
In 1975, the NRC was reviewing applications for construction permits based on preliminary BWR-6 design details, and completion of the final design details lagged significantly behind the start of construction.
Accordingly, as permitted by its regulations, the NRC issued construction permits without complete or final detailed design information.
As that information was later submitted during the operating license review, licensing problems sometimes resulted becauss some information was unsubstantiated.
The end result was increased NRC review effort in some areas.
This delay in the review process may have had an economic impact on the licensee but there was no safety sig-nificancebecausethelionsingreviewwassimplydelayed.
Status Before the first BWR 6 operating license was granted, the NRC reviewed and ap-proved detailed plant design information.
The following BWR-6 Mark III designs have been approved by the NRC:
Clinton 1, Grand Gulf 1 and 2. Perry 1, and l
River Bend.
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3.2 Amount of Marcin Between Desion Calculations for Core and Doeratino Limits hatt The Reed Report noted the following with regard to the BWR 6 core at the pre--
liminary design stage:
.(1) Design theisal margin was not sufficient to avoid power derating (a re-duction in allowed power level) to as low as 40% of the intended rated power to meet operating limits during portions of the core operating cycle.
Such a power derating would limit the reactor to operate at onl some fraction of its rated power, a substantial economic cdnsideration.y (2) Calculational models with inadequate experimental verification could have proven to be nonconservative and might require a power derating of 5% to 10%.
Safety $1onificance None.
Derating a plant to maintain acequate margin in operating limits is an economic issue, not a safety issue.
Status Today, cores are operating at or near the operating limits (not safety limits),
as design thermal margin 15 maintained, while using new fuel designs, less 4
conservative calculational models, and revised operating conditions.
This generally requires revised technical specffication operating limits for each operating cycle.
Nuclear power plant licensees are maint ing adequate safety margins in their operating plants by_ adhering to technical specification operating limits.
The need for power deraG ng is marginal, and it is g uerilly avoided by operating plants according to cycle dependent technical specifications that define the operating limit minimum critical power ratio for that operating cycle, using NRC staff-aanroved models and calculational methods.
3.3 Imp - -
Cold Shutdown Reactivity Marcin on BWR 6 Core Desian Issue The Reed Report noted that the design calculation models were inadequate to ensure that the cold shutdown reactivity margin for the BWR-6 could meet the stuck rod margin requirements in a plant's oper. equilibrium core ating license. -
Safety Sionificance None.
The concern was and is economic because clant shutdown and/or limited plant availability can result when a licensee cannot demonstrate adequate shutdown margin.
4 NUREG-1285 10 I
i
I l
t Status All operating BWRs have technical specifications that require shutdown margin 4
be maintained and that the plant be shut down if measured shutdown margin s
inadequate.
Calculational models for the final design equilibrium core will better reflect the burnup experience with cores that contain gadolinia in order to maintain a i
flatter reactivity response at core equilibrium.
3.4 Impact of (nd of Cycle (E00) Scram Reactivity insertion Rate on Core Ful1~
Power Life Issue l
5 The Reed Report noted that reduced scram response because of unfavorable void
(
coefficients and the design scram reactivity curve at E0C could require derat-ing up to 20% to meet operating limits.
Safety Sionificance None.
The concern is the economic cost of derating (reduction in allowed power level) to meet regulatory limits.
?
Status GE has addressed the economic consideration of plant derating to meet operating i
limits at EOC operation through the following' improvements:
I (1) improved fuel design (fewer negative coefficients)
(2) _ improved calculation models (3) design modifications to the BWR 6 scram system for more rapid insertion of rods (4) highly cycle-dependent (and core exposure-dependent) technical specifica-tion operating limits
,(5) recirculation pump trip provisions added to all BWR product lines 3.5 Lono-Term Effect of Radiation on Core Internals Issue The Reed. Report noted that uncertainties in estimates of radiation and corro-sion damage to BWR 6 core internals did not provide assurance of a 40 year lifetime of service.
Core internals might have to be replaced earlier to pro-vide assured structural integrity for continued operation.
Replacement of permanently installed core internals would result in substantial rr + tor Awn-time.
Also, replacing these core internals would be difficult bece t.-
to them is difficult and workers would be exposed tn high levels of
(
.t.
i i
NUREG-1285 11 7.
__7__
$4fety Sionificance
+
Two areas in the reactor internals were identified that could receive enough
- radiation fluence to significantly affect the material properties.
These ware the top guide and the mid-plane of the shroud.
Although not statbd directly in the discussion in the Reed Report, the apparent concern was that the material properties could be degraded to the point where the components could fail.
Failure of some core internals could hinder (but not prevent) shutdown of the The GE analysis indicated that there would be sufficient margin to reactor.
insert rods to achieve shutdown, even with channel interference or loss of spacing.
Status One effect of radiation on core internals and support structures that was recog-nized in the early 1980s is that austenitic stainless steel becomes susceptible to stress corrosion crackin Cracks have been found in neutron monitor guide tubes in at least six BWRs.g. Cracks have also been found in control blade handles and sheaths.
GE has evaluated the possibility of irradiation-assisted stress corrosion cracking (IASCC) in other components, some of which were mentioned in the 1975 study. The top guide and shroud are still unlikely to last the 40 year life for which the reactor is licensed, but it is believed that the core plate will not experience enough neutron fluence to be affected.
GE has been actively involved in developing non-destructive evaluation (NDE) equipment and proce-dures to detect IASCC, and in developing a methodology to justify continued,
operation with cracked components where such operation would not compromise safety.
Should the assembly become so degraded by cracking and loss of toughness that the assembly failed during a seismic event, failure could occur at several locations, and rod blockage or loss of the guide function might occur.
GE believes that the core plate is not likely to receive enough neutron fluence to become susceptible to cracking.
Nevertheless, the threshold value of fluence is not yet known with certainty, and further study of this subject is being pursued.
Although Section XI of the Boiler and Pressure Vessel Code of the American Society of Mechanical Engineers (ASME Code) requires visual inspection of core support structures every 10 years, the postulated crack locations may not be accessible for TV viewing.
GE has been actively working on a methodology to perform remote ultrasonic inspection of the suspect locations.
If this proves feasible, ice to determine whether a generic problem existsthe top guid long serv The staff believes that current monitoring, surveillance, and inspection pro-L grLes will icentify any incipient failure of core internals before failure, and that the radiation levels associated with plant operation are not likely to result in reactor safety problems from materials failure in BWR core internals.
\\
NUREG-1285 12 1
e 1
3.6 Dearte of Proof of Accuracy of Transient Desion Methods-Issue
)
The Reed Report notes that there were large calculational uncertainties because of inadequate verification of transient design methods.
This inadequate verification could lead to reduction in allowed level of power operation.
Application of more accurate methods, or reduction of these uncertainties by better verification programs, could result in smaller margins being permitted in thermal hydraulic transient analyses that are perfomed to ensure that the plant does exceed its thermal operating limit.
Safety Sionificance The concern was primarily economic with potential power derating being required to meet regulatory operating limits.
Status Better calculational methods have been daveloped and verified against plant transient tests.
In parallel with these tests, more sophisticated computer.
codes modeling the reactor core behavior have been developed.
The problem has been resolved (the resolution of Generic Issue B 19) with the staff approval and licensee implementation of more sophisticated core modeling codes, j
j 3.7 ImpactonfuelIntecrityofReducedModeratorTemoeraturedueto Equipment rallure,
[
Issue The Reed Report noted that excessive fuel failures due to pellet-cladding in-teraction (PCI) were causing power derating to reduce the leakage and ' dispersal of radioactivity into the reactor cooling water.
Prolonged overpower trans-ients due to loss of feedwater heating or other coolant temperature reduction transients, could lead to PCI failures,and challenge thermal hydraulic design r
limits.
Safety Sionificance 2
The rapid subcooling and reactivity spike resulting from loss of feedwater heaters is reflected in fuel failures induced by PCI and leads to some increase in personnel radiation exposures.
Such equipment failure and resulting fuel failure is to be avoided, and the increased exposure to trary to the ALARA (as low as is reasonably achievable) plant personnel is con-exposure reduction objectives, but the reactor safety implications are minimal beyond that.
Status The issue of fuel integrity is not a problem because it is addressed by the following measures:
(1) preconditioning of fuel during the early pha,ses of a new operating cycle (2) use of new fuel design (barrier fuel)
- NUREG-1285 13
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."WI % g.
6 %
7
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. s i
I (3) provisions for' thermal power monitor for delayed overpower trip (4) continued compliance with technical specification core operating thennal limits 3.8 Performance of Relief Valve Auomented typass (REVAB) System lssue
+
lhe Reed Report notes that the scram insertion requir6 ment for plants designed with the REVAB have not yet been achieverl.
$afety$1onificancel$tatus t
None.
This issue has no safety significance and is not relevant because the REVAB system has been deleted from the design of GE BWRs and is not used in'any l
operating BWR.
- 3. 9 Impact of Hydrodynamic Phenomena on Containment Desions i
Issue A general ' concern over the (then) current state of containment-related issues is reflected throughout the Reed Report, with. reference made to the containment issues in the Executive Summary, in the section entitled Nuclear Systems, and in the section entitled Mechanical Systems and Equipment.
In several cases the same issue is discussed in different sections but with a different perspec-tive or with emphasis on particular elements of the technical issue.
The issue of hydrodynamic phenomena and their impact on containment designs, dis-l cussed throughout the report, is identified as " Impact of Recentl Discovered i
Phenomena on Containment Des 1gns" in the 25 issues identified b E.
The i
report says:
"Because of types (Mark I, 11 and Ill) phenomena recently discovered, all BWcontainment are undergoing extensive additional analyses to evaluate structural adequacy. As a result of these anal as Mark I are likely to be redesigned and retrofitted." yses, Mark 11 as well Safety Sionificance The Reed Report reflects the uncertainty present in 1975 surrounding the dis-covery of additional containment loads created by suppression pool phenomena related to safety relief valve (SRV) air clearing, pool swell and high tempera-ture steam condensation.
These phenomena were identified durIng early testing of the Mark 111 design, which was initiated in 1973, and by the experience at two German BWR Mark I containments,in 1972.
At the German plants severe vibra-tory loads on the containment structure were experienced during ex, tended SRV operation.
In 1975, concerns also were being raised by former employees of GE, and hearings were held before the Congressional Joint Committee on Atomic Energy regsrding the impact of hydrodynamic loads on BWR containment designs.
~i The safety significance of this isst.< was that the additional loading created by these phenomena during an accident or transient could jeopardize the integrity' of the containment structure, drywell, and/or equipment and structures near the suppression pool.
Failure of the containment or drywell structures could have serious consequences'during certain reactor accidents.
NUREG 1285 14
1 Status 1
- After the Reed Report was issued, BWR owners, working with GE, completed exten-sive testing and analyt,es, resolving all technical issues related to suppression 1
- pool hydrodynamic loads.
Both generic and in plant testing weHi performed to i
provide an expanded data base on which conservative loading definitions could i
- be developed.
To reduce loads created by SRV operation, new SRV discharge cuencher designs were approved and installed.
Additionally, various plant-specific modifications were made to stren needed to restore design safety margins. gthen the containment structure asThe NRi r
issues to guide, track and document resolution of these technical concerns, as follows:
(1) Generic Issue A-6, Mark I Short Term Program.
The resolution was docu-i mented in NUREG 0408.
t (2) Generic Issue A-7, Mark I Long Term Program.
The resolution was documented a
in NUREG-0661.
l (3) Generic Issue A-8, Mark II Program.
The resolution was documented in NUREG-0487 and NUREG-0808.
(4) Generic Issue A-39, Determination of SRV Pool Dynamic Loads and Temperature Limits for BWR Containments.
The resolution was documented in NUREG-0802.-
(5) Generic Issue B 10, Behavior of BWR Mark III Containment. The resolution was documented in NUREG-0978.
3.10 Radiation Exposure from Removal of Steam Dryer / Separator Assembly j
Issue The Reed Report noted that there was a potential for significant plant person-
- nel radiation exposure from dryer / separator assembly handling for the BWR-6 Mark III design.
Safety Significance Concerns were limited to those of occupational rediation exposure.
There were
- no reactor plant safety concerns beyond the ALARA issue.
The issues' involved were primarily economic considerations associated with decreased availability due to a lack of maintainability, and the ALARA issue of maintaining occupa-tional exposure to low levels.
Status After the Reed Report was issued, the BWR-6 Mark III design was modified to allow underwater transfer of the dryer / separator assembly, thereby reducing occupational exposure rates, particularly during refueling.
The NRC staff considers this modification an excellent example of field feedback, self-analysis, and implementation of ALARA guidelines.
~
NUREG-1285 15 f
- T T-g
..h-84
4 p*&,-=
g
3.11 Level of Testino of Marh !!! Containment Imt The Reed Report in sev'eral sections reflects concerns over the adequacy of testing planned to investigate suppression pool phenomena for the Mark III com tainment. Although this concern is related to the general issue of suppression pool hydrodynamic loads, it it specifically related to questions over scaling of Mark III tests to determine pool swell loads resulting from a loss-of-coolantaccident(LOCA).
The concern stems from initial Mark !!! tests that were conducted with non uniform scaling; a full-reale sector of the suppression pool was simulated while the drywell and boiler simulation was 1/3 scale.
The
' Nuclear Systems section of the Reed Report recommended that " full scale" boiler and drywell tests be performed along with consistent 1/3-scale tests.
In the J
Mechanical Systems and Equipment section of the Reed Report,lete i
the recommendation is conditional; it recommends that 1/3 scale testing be comp i
as possible, and expanded, if necessary, to resolve uncertainties.
1 Safety Sinnificance
)
The safety significance of this issue deals with the uncertainty over load L
definition for suppression pool phenomena.
If the test data used to define L
loads were based on improperly scaled test models, then by extension the load definition used to evaluate containment structural response would be inadequate.
Status Pool swell tests were continued for approximately 4 years after the issuance of.the Reed Report.
Testing was conducted on a variety of scales and configura-tions in order to confirm the use of conservative scaling factors in load
.'l definition.
A full-scale model of the drywell,-boiler, and suppression pool was not needed.
The GE technical resolution was documented in a series of reports, NEPT-13377 20550, 21853, 13407, 13426, 13435 21596, 24648, and 24720.
The loaddefinItionreportfortheMarkIIIcortaInments(GEdocument22A7007, February 25,1982) was reviewed and approved by the NRC.
The NRC also initiated Generic Issue B-10. " Behavior of BWR Mark III Containment," to address this issue; NRC evaluation and resolution of this generic issue was addressed o
in NUREG 0978 (August 1984), which documented the NRC, staff acceptance of modifications and results of the load definition report on the Mark Ill containment.
3.12 Presence of Detectable Plutonium Inside the BWR Turbine 5
Issue The Reed Report noted that detectable amounts of plutonium produced by trans-mutation of uranium had migrated beyond the fuel pin boundaries and deposited inside the turbine of BWR reactors.
Safety Significance Plutonium is a source of long lived alpha radiation, chemically related to i
. calcium. When it is ingested it tends to deposit in the bone This subjects the tissue to long-term ionizing radiation, which can produce cancer.
2 NUREG-1285 16
+
5tatus
+-
Trace amounts of plutonium deposited inside the turbine are carried by steam from the reactor core to the turbine.
The plutonium can be produced from tramp uranium, which are trace amounts deposited outside the fuel pins, or f. rom leak-ing fuel pins.
Experience has shown that essentially all of the plutonium
' formed in the fuel stays there.
Further, analyses of reactor water show that 1
, the plutei= nntent is typically less than 1% of the permissible drinking j
water level. These trace quantities are removed by the rasetor water purifica-tion system.
Plutonium contamination in BWR turbines is not a significant problem.
l 3.13 The Effect of $1oshino of the Suppression. Pool on Mark III Steel Containment 5tructure Desion 4
Issues i
The Reed Report noted that testing asociated with Mark III containment was incomplete and the potential for dynamic buckling resulting from seismic sloshing of the suppression pool had to be considered in the design of the steel containment.
Safety Significance i
Buckling of the steel containment shell from sloshing of the suppression pool in a seismic event may result in failure of the containment functional capability.
{
L
$tatus i
The potential for buckling of the steel contafnment shell as a result of slosh-ing of the suppression pool is being handled in several different ways.'
At Perry and River Bend, the annulus between the steel shell and shield building is filled with concrete up to a level above the suppression pool.
Through analy-i sis, it has been demonstrated that seismic sloshing of the pool theh cannot result in buckling of the steel shell.
Thus, the containment functional capability cannot then be compromised, and there is no safety significance.
At the design stage, buckling of the steel shell without concrete backing was considered in the Perry and River Bend plants, and was reviewed by the NRC staff.
The design was found to have met the staff's buckling criteria.
In the case of Grand Gulf and Clinton, the containment structure is not a steel shell, but is concrete, not subject to potential buckling from seismic sloshing.
j Buckling of steel containment shells, including consideratioh of dynamic t
responses of the she11r was studied at Lockheed Palo Alto Research Laboratory under contract with NRC.
The staff's buckling criteria are based mainly on the results of this study (NUREG/CR-2836).
3.14 Evaluation of Fuel Transfer Accident in Mark III Containment Issue The. Reed Report noted that the potential for a fuel transfer accident in the i
Mark III containment had not been evaluated.
NUREG-1285 17 e
^
4 t
i'i-In 1975, GE had not completed the desi n of the Mark I!! containment. This containment.was similar to the pressur zed water reactor (PWR)-style contain-
.sent where the spent fuel storage facility is located outside of the reactor
' building and away from the refueling floor.
In the Mark III design the spent fuel pool is located at a lower elevation than the refueling floor,whereas in the PWR designs the refueling floor and the fuel handling floor in,the fuel building are at the same elevation.
A cone.orn was raised that spent fuel would have to be transported in the Mark III containment from the refueling floor l
elevation to the lower fuel building elevation.
Since this spent fuel had recently been in the core, it would have. a high rate of decay heat generation.
2 If the fuel were to become immobile during the fuel transfer process, there might not be adequate cooling for the fuel bundle, and the radiation shine through the surrounding walls might create a new and different-type hazard to plant personnel.
In addition, an elaborate valving arrangement was needed to preventthewaterintheupperpool(insidecontainment)fromdrainfngdown_
into the spent fuel pool.
i Safety Sionificance
]
+
The potential safety significance of these postulated accidents is centered around two areas:
radiation exposure considerations and the potential breaching of containment.
The stuck fuel bundle in the transfer mechanism could represent a radiation expnsure concern for. workers in areas adjacent to i
the fuel transfer tube and for those on the refueling floor from gas being-released from fuel bundles as they heat up because available cooling is not i
adequate.
m The' simultaneous opening of both transfer isolation valves (one at the refuel-n ing floor in the. reactor building and the second in the fuel building-in the spent fuel pool) could breach containment and drain the u l=
flooding the spent fuel pool and the fuel handling floor.pper containment pool, If.a spent fuel bundle were to be stuck in the transfer tube at the time of the valve failures, the bundle would' overheat once the upper pool was drained; this would result in a release of radioactivit to the containment atmosphere, resulting in increased exposure to the fuel hand ing personnsi in the vicinity.
Status Since the Reed Report was issued in 1975, GE has completed an evaluation of i
these potential accidents.
In addition the NRC staff reviews the potential fuelhandlingaccidentaspartoftheIIcensingprocess.
In the GE design quate protective measures are taken to prevent personnel from having access, ade-to areas near the transfer tube, especially during fuel transfer operations.
The NRC staff has reviewed the fuel transfer system to verify that no single-fail-ure could result in a fuel handling accident, and that all aspects of the sys-tem have the appropriate alarms and interlocks.
1 As part of this failure modes i
and effects analysis tube isolation valves,.the potential for inadvertent opening of both transfer j
simultaneously was given special attention to ensure that containment will not be breached and that the upper containment pool will not be drained.
Thus, the concerns raised in the Reed Report have been satisfac-torily addressed to ensure that the use of the inclined fuel transfer system will not result in any significant increase in the risk to the health and safety of the public or to plant personnel.
NUREG-1285 18
_ -._ - ___ ____.-_.+_, - _ _
i i
u E.*
3.15 Isonet of Core Desion and Licensino Criteria on SWR Capacity
!ssue i
-The Reed Report contained a table that identified several. potential problems, some having safety significance, that could affect plant availability and capa-l city factor.
Safety Sionificance
(
t The concern was primarily economic, with shutdowns and power derating resulting from either equipment probleas' or from a licensee's inability to meet regulatory requirements.
+.
Status i
/
These problems have been resolved through the following:
i (1)~ Fuel densification problems were resolved by changes in fuel design.
(2) Emergency core cooling system criteria in Appendix K of Title 10 of the Code of Federal Regulations Part 50 (10 CFR 50) have been satisfied.
t (3) Channel box wear and cracking was caused by flow induced vibration of incere instrument and startup source tubes.
The problem was resolved by r
eliminating bypass flow holes in the lower core plate and adding two holes a
l in the lower tie plate of each assembly to provide an alternate flow path.
j j
. (See also discussions in Sections 3.5 and 3.18 on channel box problems.)
All other problems listed affecting plant availability and capacity factor are identified and addressed elsewhere in this evaluation.
i 3.16 Adeouacy of Design Procedures To Ensure Compliance with Licensino criteria Issue t
The Reed Report raised the following concerns regarding quality assurance (QA)
_for the BWR-5:
(1) GE had no identifiable systems engineering organization to provide independent evaluations of BWR designs at critical points in the program.
(2) GE's existing procedures for BWR systems design reviews needed improve-ment, and additional procedures were needed for QA for the BWR-6.
Safety Sionificance There was a lack of confidence that applicable licensing requirements would be
[
implemented and documented.
4
- J NUREG-1285 19 b
s
-..m.-m-
'_4 Status l
t At the time the Reed Report was issued, the GE nuclear QA program for the SWR 6 had not been completed.
Since then, a program has been completed that complies with the applicable NRC requirements, codes, and standards, and the NRC has given its approval for operating license applicants to reference this QA program in the Final Safety Analysis Report for a plant.
This GE report (NED0 11209 03A and 04A, currently approved by NRC staff through Revision 6. dated July 1986) describes the approved QA program for design, fabrication, and procurement activities involving safety and safety-related structures, systems, and com-ponents of GE nuclear power plants.
3.17 Consistency of Decree of Verification of Calculational Models Issue The Reed Report raised a concern that calculati nel models were not thoroughly reviewed and verified by comparison to experimental data to ensure adequacy.
Safety Sionificance A calculational model that is not adequately verified by comparison to results using experimental data can lead to nonconservative errors in results, and uncertainty in operating limits derived from reactor safety analyses.
Status GE has completed major experimental programs for verification of currently approved models, and verification problems have been resolved.
The NRC-staff has reviewed and approved all calculational models that are necessary to be used in licensing of operating BWR plants.
3.18 Possibility of Control Rod Binding Due to Fuel Channel Creep Issue The Reed Report noted that fuel channel life was projected to be 8 to 10 ye6rs (two complete refueling cycles) rather than the desired 15 years, due to thermal creep and control rod binding.
Safety Sionificance Binding of control rods can cause slower negative reactivity addition, thereby invalidating the licensing assumptions and increasing the severity and conse-quences of transients and accidents.
Status Today, fuel channel shuffling requirements and scram-time testing technical specifications ensure against degradation in scram time.
The NRC staff has approved channel surveillance programs, in conjunction with relocation and rotation to minimize irradiation-induced channel bow, and spe-ciel rod motion testing for core cells exceeding core residence program guide-lines as ways to extend channel lifetime.
NUREG-1285 20 l
1
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- m=-
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t
-3.19 Compliance of Desich Work and Reviews with Written Procedures 4
1 i
Issue i
The' Reed Report identifies the following concerns regarding the BWR 6 Mark III that arose from findings of a GE internal audit:
'(1) Design reviews, internal procedures, and QA audits were not always con-ducted in conformance with established written procedures.
(2) QA audits conducted by GE revealed instances of nonconformance with BWR Systems Department engineering practice and procedures.
(3) Staffing and organization of design assurance efforts in the BWR Systems Department did not optimize its effectiveness in departmental activities.
t' R
(4) There was a lack of coordination between the procedures and QA (P and QA)
- i organization and GE components audited.
Safety Sionificance A proper internal audit program is needed to ensure that inadequacies in proce-l dures and noncompliance with procedural requirements will be discovered and corrected.
Status
. As described previously,*the GE QA program has been reviewed by the NRC staff,
~
and GE now has an effective internal audit program,_a part of the GE Quality Assurance system.
GE audit reports are available to and inspected by the NRC.
Experience has demonstrated that the GE program is effective in finding devia-tions and deficiencies, as it was designed to do.
3.20 Absence of Availability Goals in Design Procedures Issue The Reed Report discusses instances of nonconformance with GE procedures in-i volving issues that are basic to the achievement of design integrity and that affact plant availability.
In particular the study was concerned with achiev-ing an optimal balance in the engineering, design goals between availability and safety.
The study noted in particular that many design procedures did not have
, availability goals.
Safety Sionificance The absence of availability goals design integrity.
Regardingavailabilitygoalsby itself, has no impact on safety-related in its licentin NRC uses safety design requirements as found in,its regulations,g reviews the its Standard Review Plan (NUREG-0800), its Regulatory Guidelines, and other MC. position papers, rather than availability goals.
4 NUREG-1285 21
~
gs, l
-Although the NRC has not established quantitative availability requirements'for safety systems, unit availability can be limited by technical. specifications,
and require shutdown when key safety systems are unavail-j
.that prevent start able.
All operatin SWRs have such technical specifications, and plant avail-1 ability can be affected by these technical specification limits.
)
- 3.21 Seismic Caoabilities of 8 x 8 Fuel Soacer g
!ssue 1
The Reed Report raised a concern related to the seismic capability of spacer i
- design for 8 x 8 fuel.
Specifically, potential loss of core coolability be-l cause of fuel spacer failure under the combined loading of an earthquake and a loss of-coolant accident (LOCA) was envisioned as a possible impediment to l
licensability.
I l
Safety Sionificance
'l Maintaining the core in'a coolable' geometry during seismic event helps limit
~
the consecuences of a postulated LOCA to acceptable release levels.
. Status GEhascompletedtheseismictestin$ofthefuelassemb1PA,datedOctoberf98 The NRC staff i
spacer and has reported the results in NEDE 21175-has reviewed those results and accepted the design for'use in SWR cores.
3.22 Extent of Life of Position sensor in Traversino In-Core Probe System Issue.
The Reed Report addresses operational problems with the traveling in-tere probe
.(TIP) system,. including bending and contamination of the guide tubes.
Safety Sionificance Technical specifications and plant procedures require periodic calibration of local power range monitors that input to reactor protection systems using the TIP system.
Power distribution information obtained from the TIP system is used to maintain core operating limits.
Unavailability of the TIP system would
-t prevent plar,t operators from obtaining certain information necessary for starting up the plant.
Unavailability of the TIP system could then adversely L
affect plant availability.
Status l
Service experience with modified TIP systems designed for better availability i
demonstrees that longer life and improved accuracy (compared with earlier models) is being achieved.
Efforts to further improve the operational useful-ness and dependability of the TIP system are ongoing.
NUREG-1285 22
.. - ~
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j
'3.23 Radiation Levels Outside Biolonical Shield and Drywell Issue-t The Reed Report noted that unexpected and excessively high levels of radiation outside the biological shield and/or drywell containment would constitute an i
occupational radiation exposure problem.
Safety Sionificance s
Imperftetions in shielding design can result in unexpected radiation streaming through unrecognized pathways.
This is a personnel radiation exposure problem.
It also creates difficulties in maintaining and servicing affected parts of the plant when radiation levels are high.
1 Status High levels of shine radiation were observed during startup, particularly in early plants.
However, this is no longer a problem in operating plants.
To-i prevent such occurrences, it is standard practice to perform startup radio-logical surveys to confirm radiation levels and to identify unexpected ones.
Licensees have identified all such pathways by actual surveys and have elimi-i nated them.
Such programs ensure that radiation exposure levels for workers do not exceed NRC established limits and conform to ALARA guidelines.
'3.24 Stress Corrosion Cracking in Dresden 1 Control Rods
[
Issue The Reed Report.noted that control rod lifetimes might be limited because of stress corrosion cracking in the control rod blades.
This could lead to problems of (1) limited control rod life (2) loss of reactivity worth (leaching o) absorbermaterial)
(3) continued operability (cracking of sheath)
Safety Sionificance There is a potential for reducing the shutdown margin to below that required by technical specifications.
Status GE has performed an analysis of the safety implications of control rod cracking and consequent loss of rod worth. The results show that any loss of reactivity worth would be revealed b a shutdown margin test before the loss could jeopar-dize safe shutdown capabi ity of the reactor.
In addition (1) Problems with control rod blades identified through operating experience were resolved by licensee actions in response to NRC IE Bulletin 79-26, i
Rev. 1, " Boron Loss from BWR Control Rod Blades," dated August 28, 1980.
J NUREG 1285 23 i
1 W
4.
(2) Later problems involving cracking of advanced design blades in the sheath at:the handle region have been evaluated and are being addressed by a continuing surve'11ance program, t
(3) Improved hybbid-hafnium control rod designs and better control of water chemistry have alleviated, but not eliminated, the problem of control rod blade degradation with use.
't L
The broader issue of stress corrosion cracking in stainless steel piping asso-ciated with nuclear reactors is addressed in Section 4.6 of this report.
3,25 Peak Pressures in ATWS Calculations for BWR 3 Plants Issue i
- The Reed Report noted a potential for damage to the reactor vessel dye.to pos-sible :eak pressures of 1600 to 1650 psig during certain postulated events for the BWR-3, particularly the anticipated transient without scram (ATWS) event.
Safety $1onificanes i
l Overpressurization and failure of a reactor vessel would result in consequences i
beyond those acceptable for licensing a nuclear power plant.
Status Such pressures.resulting from a transient event could occur only at elevated temperatures when the pressure vessel material is in a ductile state and is thus less subject to damage by an overpressure event.
Further, more refined calculations by GE using better anal tical methods demonstrate that peak pres-
".sures in such an event would be far ess than the 1600 to 1650 psig estimated in 1975..
Interim resolution of the ATWS. issue was provided by improved procedures and operatortrainingIonpumptrip).and through implementation of certain hardware L
(e.g.,recirculat The ATWS issue was finally resolved when i
NRC issued the ATWS rule (Title 10 of the Code of Federal Re ulations, Section l
i 50.62 (10 CFR 50.62)), in July 1984.
In response to this ru e, plant-specific measures, including hardware modifications, have been made in all operating BWR plants, and further modifications will be made in some plants.
In October e
1986, the Nr0 accepted the GE licensing topical report NEDE-31096 P "Antici-pated Transients Without Scram means that licensees may now re;ference this repo.rt in their plant-speci i
actions.
f i
l;
(
.NUREG 1285 24 t
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4 OTHER SAFETY-SIGNIFICANT ISSUES 1DENTIFIED BY THE NRC STAFF IN'THE REED REPORT Through its most recent review and evaluation of the Reed Report, the NRC staff
'i identified several safety-significant issues in the report that had not been highlighted by either the NRC staff or by GE in its 1978 status report on the
.l Reed Report.
These are identified and discussed below.
J None of these issues involve any safety consideration not already identified and appropriately addressed by the staff.
4.1 Combination of LOCA Induced Loads and Safety Relief Valve (SRV) Actuation i
Loads for Mark III Containments j
l3 4
Issue stems and Equipment, The Reed Report, in the section entitled Mechanical.Sy/ licensees to consider l
cites a concern that the NRC might require applicants combined LOCA-induced hydrodynamic loads and SRV loads in the evaluation of suppression pool loading phenomena and design of the Mark 111 containment.
The report further notes that it is not unreasonable to postulate SRV opera-tion concurrent with a LOCA.
The.GE status report did not explicitly identify this issue. This issue could, i
I L
however, be considered a component of the overall issue of hydrodynamic phe-nomena identified and discussed previously.
The Reed Report recommended that a high priority be assigned to the resolution of this issue, and that conserva-tive containment design loads should be used by architect / engineers in the design and construction of plants.
This approach was suggested to minimize the-likelihood that future redesign or plant modifications would be needed after
' testing and the NRC review were completed.
Safety Significance
.The safety significance of this issue, as acknowledged in the Reed Report, is that the combination of LOCA and SRV loads could result in a. higher total loading condition. The larger loads could threaten the integrity of tt.e containment structure under accident conditions, or could reduce the safety margins in the design.
Status NRC now requires applicants / licensees to consider the combination of SRV and LOCA suppression pool loads; however, the NRC has evaluated and approved the GE methodology for the combination of these hydrod NUREG-0798 docu-ments resolution of this issue for the Mark !!!ynamic loads.
containments as part of the resolution of Generic Issue B-10, " Behavior of BWR Mark III Containment"; reso-lution of this issue for the Mark I and Mark II containments was documented as part of the resolution of Generic Issues A-6, A-7, and A-8.
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~ 4.2 Jet Imoincoment on the Weir / Pool in a BWR Mark'III Containment l
Issue.
The Reed Report, in the section entitled Mechanical Systems and Equipment, included the following recommendation:
"The possibility of a direct pipe break jet impingement on the weir / pool and its asymmetrical effects should be ex-amined.
Preliminary judgement is that this is not serious." - The NRC staff was unable to locate any other clarifying information on this issue in the report.
This issue was not identified or discussed in the GE status report provided in 1978..
Safety Sinnificance If direct pipe break jet impingement on the weir / pool were to occur, the jet impingement loads could cause structural failure of the weir wall.
Failure of the weir wall in the extreme could cause an uncovery of the suppression pool vents which, in turn would lead to bypass of the suppression pool.
For t
certain accidents, significant steam bypass of the suporession pool could re-sult in overpressure failure of the containment.
If the asymmetric suppres-sion pool loads on the weir wall were sufficiently large, they would have the same consequences.
Status-Jet impingement effects resulting from postulated pipe breaks are not un!que to BWR Mark III containments and are addressed for all plants during the course of licensing review.- The general consideration of jet impingement loads on struc-tures and equipment includes those effects, if any, on the weir wall in a Mark III containment.
For asymmetric suppression pool loads, the effects of such loads on the weir wall is minimal, because they are bounded by'other weir wall loads (e.g., chugging load, depressurization load).
Asymmetric pool swell loads were addressed in NUREG-0978, in the resolution of Generic Issue B-10,
" Behavior. of BWR Mark III Containment."
4.3 Main Steam Isolation Valve Leak Tichtness Issue The issue of leak tightness of main steam isolation valves (MSIVs) was identi-fied in the Reed Report in the section on Mechanical Systems and Equipment, but I
was not discussed in the GE status report provided in 1978.
i Main steam isolation valves (MSIVs) have been notorious for leaking at high rates when they are tested during the 18-month leak tightness testing that is generally required by the technical specifications.
Most plants have a tech-nical specification leak rate limit of 11.5 standard cubic feet per hour (scfh) per valve.
At some plants the as-found leak rate has been as high as 4500 scfh.
-With such high leak rates, the MSIV-leakage control system (MSIV-LCS) probably would not be capable of performing its safety-related function of removing the leakage from between the closed MSIVs following a design-basis LOCA.
NUREG-1285 26 j
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.l Safety Sionificance n
In its evaluation of the safety features of nuclear power plants, the past prac-tice of the staff has been to give no credit for any structure, system, or component that was not safety related (sometimes referred to as safety grade).
i l
Given this past practice, following a design-basis LOCA with no credit or non safety-related coisponents, and assuming the single failure of one MSIV to
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close, the design basis maximum allowable leakage through the MSIVs, for most plants, is 11.5 scfh.
This limit on MSIV leakage is to maintain the offsite o
radiological consequences to within a small fraction of regulatory limits in the event of an accident.
Thus, if the MSIVs were to leak at a rate greater than I
11.5 scfh, and particularly at a rate that caused the MSIV-LCS to fail, tha offsite consequences could exceed the regulatory limits in the event of a
.i severe accident.
-Status In recognition of this continuing problem of MSIV leakage, and the potential l
consequences in terms of offsite doses, the NRC staff early initiated Generic Issue C-8, "MSIV Leakage and Leakage Control Systems Failures." This generic issue considered the actual natural phenomena associated with the behavior and the characteristics of radion'etive materials and the historical capability of "nonsafety related" components to survive seismic events.
In assessing the consequences of MSIV leakages, credit was given for fission product decay plate-outoncoldsurfaces,andgravitationalsettling,andforarealistIc evaluation of the actual materials that would be transported along the main i
steam line.
Because it is assumed in design basis accient analys'es that offsite power will be lost following a LOCA (as a result of the tripping of the turbine generator
+
and failure of offsite power), no credit was given for any equipment that was not powered from the emergency diesel generator busses.
The analysis performed t
under Generic Issue C-8 indicates that the leak rate through MSIVs could be as high as 1500 scfh without using the MSIV-LCS, and the offsite doses would be less than those specified in the regulations.
The study identified a method of calculating this leakage rate, but the actual leak rate would have to be 4
determined on a plant-by plant basis. This information was documented in NUREG-1169, published in August 1986.
.MSIV leak tightness was a concern in 1975, and it is still a concern that has i
not been fully resolved. The BWR Owners Group (BWROG) formed a committee to evaluate this same issue independently, with GE giving technical support to the BWROG committee.
This committee generally found that the high leakage rates were attributable to valve maintenance practices.
For those plants that have adopted the BWROG recommendations resulting from their evaluation, the as-found MSIV leak rates have generally been within the plant-specific technical speci-fication limit, or within a factor of 2 or 3 of that limit.
For example, Peach Bottom 3, had typical as-found leak rates of over 3000 scfh for each of the MSIVs.
After following the BWROG recommendations, the next as-found leak rates were found to be less than 11.5 scfh for seven of the eight MSIVs and approxi-mately 14,7 sr.fh for the eighth MSIV. This demonstrates that the MSIVs can be maintained within the w respective technical specification leakage limits, and that the use of the leakage control system is not necessarily the optimum method for handling the leakage through the MSIVs in the event of a LOCA.
NUREG-1285 27
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The technical specification limit of MSIV leakage is conservatively set to 4
ensure that offsite dose consequences of a main steam line break are a small i
fraction of the regulatory limits in 10 CFR Part 100.
Although MSIV leakage is.
en issue.of continuing concern, ion of SWR plants as the MSIV leakage issue the current state of the art and conservative limits justify continued operat pursued.
a 4.4. Control of Design of Purchased Comc'onents.
Issue 5
-The Reed Report identifies concerns that i
(1) Because GE's Nuclear Engineering Department (NED) relies almost entirely on other vendors' design expertise to produce components to purchase.
I specifications, GE needed to develop more engineering competence and design expertise in hardware purchased from vendors,lves, flow control particularly valves (e.g., main steam isolation valves, safety relief va
?
valves,etc.).
(2) GE needed to implement a procurement policy that provides for engineering reviews and approval of design details for materials of critical compo-t nents that are purchased from vendors.
3 Safety Significance I,
The failure of purchased components used in GE safety systems or in systems im-1 portant.to safety could prevent those systems from performing their intended-j
,L, functions.
Status.
I Currently, purchased components used in GE nuclear systems are appropriately-l consideredintheGEQAprogram.
(See also Sections 3.16 and 5.6 of this report.)-
4.5 Flow Induced Vibration of Jet Pumps I
Issue The Reed Report raises a concern that inherently high excitation due to tur-bulence in the-upper end of jet pumps could lead to mechanical failures caused by flow-induced vibration.
Sefety Sionificance l'
l-l-
Jet pump mechanical failures could invalidate the licensing basis LOCA analyses
-- through a failure to maintain the assumed vessel water levJ at the top of jet 1
pumps during reflood.
l-Status Subsequently, tests performed by GE demonstrated that major structural compo-nents-should withstand anticipated vibratory stress levels.
However, operating 1
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. experience revealed a problem with the helddown beams which cracked in some operating reactors.
The problem was addressed by design channes to the hold-down beams and by appropriate surveillance programs and techn' cal specification surveillance requirements to monitor jet pump operability.
These recommenda-tions and requirements were in NRC IE Bulletin 80-07, "8WR Jet Pump Assembly Failure," dated April 4, 1980; they included the use of improved holddown beam i
bars and a required surveillance program to anticipate incipient beam bar fail-Jure that could result in displacement of the jet pump assembly.
4.6' Stress Corrosion Cracking in Stainless Steel Pioino Issue The' Reed Report notes that stress corrosion cracking (SCC)_ has occurred in type 304 stainless steel piping in'several operatin occurred in nitrided stainless steel parts, furnaceg BWRs and that SCC has in bolts that have been heavily cold-worked.
sensitized components, and The Reed Report recommer.ded that GE develop replacement materials, e; pand i
studies on materials, expand study on stress levels, increase effects on en-vironmental effects. on fatigue for water chemistr relationships between operating practices and cra>ck..ig.7entrol, and study the t
Safety Sionificance Several studies have shown that pipe cracking has minor safety significance.
Both experience and analyses have shown that cracks in pipes caused by stress corrosion cracking will develop readily detected leaks before cracking develops to the point that cosplete pipe, failure will. occur.
Nevertheless, the NRC staff has determined ~ that reliance on this leak-before-break behavior is not
. suf fic.ient.-
~
Appropriate remedial reasures -- including augmented inspections to detect cracking-in.early stages -- and corrective actions are required where appropriate (see NRC Generic Letter 84-11, dated April 19,1984).
Status Since 1975, extensive cracking has been discovered in stainless steel piping in 1
BWRs.
The NRC has established two Pipe Crack Task Groups and implemented their recommendations.
The industry also has mounted an extensive effort to
' address the problem and develop remedies.
As a result of cracking observed in large and small stainless steel pipes in recent years, all operating BWRs
~
having susceptible piping have implemented an NRC staff prescribed surveillance program, with staff-approved pipe repair or replacement where appropriate.
Currently, a comprehensive set of guidelines that prov' ides the NRC positions on actions to control pipe cracking in BWRs is under development.
The NRC staff has prepared a generic letter, together with a technical report (NURE0 0313,
.Rev. 2. " Technical Report on Material-Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping"), that will be issued shortly.
This letter and report set forth the actions that plant owners must take to keep their plants.in conformance with NRC requirements related to piping integrity.
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ing, and ere in almost complete conformance with.the proposed NRC guidelines.
-e lf, in accordance with:the-forthcoming generic' letter,: individual welds are.
'found to:be not in conformance with the materials and prccess guidelines, aug-mented inspections will-be required to ensure the continued integrity of t the piping.;
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5$THE'GENF.RALELECTRICSUBTASKGROUPREPORTS.
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i This section contains the NRC' staff evaluation of each of the subtask reports l
that were prepared as input to the Reed, Report.
The reports address the
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Efollowing topics:.
.'Subtes k Topic-y j
. A: -
Nuclear Systems u
8:
Fuel-i.
C:
Electrical, Control, and Instrumentation
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. D:
Mechanical Systems and Equipment 1
E:
Materials,' Processes, and Chemistr>
.F:-
Production, Procurement, and Construction i
" G:l
- Quality Control Systems Overview 1H:;
' Management /Infornation System L@ '
I:
Regulatory Considerations
-J:
Scope and Standardization jQ
'In its own evaluation of these reports, the NRC staff'has attempted to identify l
'any issues having safety significance, and to indicate the status of-the, issue-so-far~as the NRC staff is concerned.
The staff found no issues of safety.
significance that have not already been addressed by NRC staff initiatives, with 1
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ithe possible exception of a plant auxiliary power systems: issue identified in
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the Subtask C report (Section 5.3).
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[ 5.1 ' Subtask A: Report on Nuclear Systems q
(INTRODUCTION l
W The subtask report on nuclear systems deals.primarily with.several issue's ex-uO
- pected to necessitate reducing the-allowed power level of reactors-(power a
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- derating) during portions of the core operating cycle.
These. issues-stemmed i f' Lfrom a marketing strategy that required GE to commit to:desi ns of increasing
' size and performance before the designs were adequately ver1 ied via test-data MP and field experience.
Additionally the advanced designs-were standardized on 1
- the basis of earlier designs before sufficient field experience feedback could y
- be considered; The GE task force was concernest that reliabilit / availability
! considerations would be major factors in futures procurement-eva untions by the i
e
,1 utilities,.and that field experience with BWRs, especially with'the BWR-6, would not reflect favorably on the product-
SUMMARY
OF ISSUES o
m
.Most of the issues involving systems aspects of BWR NSSS design that were perceived as contributors to power deratin in the 1975 study are addressed 1
N in the the Reed Report. The safety signif cance and current status of.the I
following Subtask A issues are discussed in Section 3 of this report in the W-listed subsection:
Issue Subsection Amount of Margin Be+ 3en Design Calcu-3.2 "g
lations for Core and Operating Limits NUREG-1285 31 h
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Impactif:ColdShutdownMarginon 3.3{
a at BWR 6 Core Design a
'IkactofEOCScramReactivityInsertion1 5
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Rate on Core Full Powen Life
- Degree of Proof of Accuracy of Transient 3.6
, Design Methods' Impact on' Fuel Integrity of Reduced Moder--
3.7
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.atorzTemperature due to Equipment Failure I
'.3 mpact of Core Design and Licensing Cri-3.15
.teria_on BWR Capacity (New ECCS Criteria).
l
- Consistency of Degree of Verification 3.17
-of Calculational Models-s Radiation Exposure from Removal of Steam 3.10 Dryer / Separator Assembly
.In its'~ detailed review of the subtask report, the NRC staff identified several subissues that are presented in more detail or-in a different context from the-o discussion of the above issues in the Reed Report.
A discussion of these addi-tional issues which impact plant availability, and their safety significance, foll ows.'.
Regulatory BEckfit Issues:
Sixteen issues expected to require _backfit to plants under con-a>
struction were identified.
c Safety Sionificance:
Some-backfit issues identified were necessary to meett
,new regulatory requirements, and some were not.
Status:-Changes were implemented where appropriate.
Incomplet,e Desian Issue:
Reload cores and behavior of equilibrium cores were not factored into the-design process for the early BWR-2 to -5 designs.
Transient characteristics of BWR-2 to -6 designs were not assessed until'after the core and circulating systems designs were frozen for hardware procurement.
Seismic design analyses were performed after hardware layout was complete,.
and the level of effort was insufficient to complete the design properly.
o Safety Significance:
The economic penalty of the failure to show design margin to operating limits in frozen designs and in reload cores creates unoue pressure to compensate for design shortcomings via the application of nonconservative and unverified calculational methods, which could re-J sult in violation of fuel integrity or LOCA operating limits.
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Reactors were licensed based on the'results of safety analyses; Am
.using NRC' reviewed and approved calculational. methods.
The' regulations 9
require that reload core designs involving unreviewed safety j
bytheNRCstaffprior.toimplementatIo.coreoperatinglimits technical specification changes (e.g.
be. approved
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. Uncertainties in Reactor Core'Desion Methods t
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Issue:
The design thermal: margin to operating limits was found to be-s significantly less~than that predicted based on field measurements, showing 1
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discrepancies between predicted and measured void reactivity worth and a 10% underprediction of the depletion rate of gadolinium rods.
The 4
25% margin provided inLinitial transient design analysis eroded to 10% by-3 the void model error, and additional uncertain 1.ies that could further-p erode thermal margin were identified.
f Safety Sionificance:' These reactor core design models are used to estab-1
- E 1ish= technical specification operating limits for fuel. integrity and LOCAs and to evaluate the consequences of transients and accidents.
Status:
Improved' calculation.models have been developed and. verified.using experimental data and plant transient tests.
These sedels have been re-viewed and approved by the NRC staff and were used ir the final safety
[
analyses (and for' reload core designs where' appropriate) for most opera-E,,
ting BWRs.
Where' uncertainties exist in these methoc1s-NRC requires that L"
they be quantified and applied conservatively in the IIcensing safety -
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-analyses and, in'some cases where pre-operational verification is not L
feasible, requires the licensee to W m confirmatory verification.
tinglimits(notsafetylimits)
Reactors are operating at'or near a
during much of the core operating b <
9odeling.to minimize the use of Utensive power derating'has been avoided via new fuel designs, a4 bounding safety: analyses, and detaileu analyses of reload cores-to ensure that core management schemes and fuel-cycle-dependent technical.specifica-tions provide maximum operating flexibility.
- License'es must maintain' adequate safety margins by adhering to technical specification operating limits.
Void Coefficient / Relief Valves Issue:.The void coefficient used in BWR transient design resulted in reactivity addition following an_ isolation (turbine generator trip) that
,was too small by a factor of 4.3 for BWR-6 equilibrium ccres as a result of changes in reactor characteristics and more realistic modeling.
Design
. scram reactivity is reduced by a factor of 5 for the E0C equilibrium core due to the high_ reactivity in voids.
Protection against overpressure o
transients of greater severity is provided by additions of relief valves, L
trip circuitry, and fast scram drive blades.
There was concern that in-crease in the number of pressure relief valves and the number of chal-1enges to these valves would significantly increase plant unavailability.
Safety Sionificance:
Greatsr relicnce is placed on safety relief valve performance to protect against ovarpressure transients that challenge pressure limits on the vestel r.d thermal limits on the fuel.
NUREG-1285 33
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, Status:z 20ther designLchangest-- such as less. negative fuel: void coef
..2 cients. the fast scram drive on SWR 6s, and recirculaticn pump lsions,- in con.ju:nction with improved scram calculation:nodels
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'duced.the severity of the transient.. There is no noticeable increase in
..have' re-L
. plant unavailability due to pressure relief transients.
A.,
Flow Control Ranoe 08 y
Issue:
The operating f. low control range was reduced for'BWRs of higher core power density; for BWR-S the nominal range was 75% to'100% versus 50% to 100% in earlier.BWR-3 designs. The reduction in range was neces-a
. factor) for equilibrium cores at EOC.sary to neettthe. design stabilit hl
" Safety S hnificance:
TTitxibili",y and requires more frequent' control rod movemen e
1 to increase' fuel failures.
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' Status:
Fuel design improvements have reduced susceptibility.to PCI fail-e9
'ure related.to control roa sovement.
w The resolution of-Generic Issue B-19.-
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" Thermal 1 Hydraulic Stability," per,mits plants to operate at higher stabilit
/
tdecay ratios,which permits removal of the design restriction on flow
= control range,.:
W CONCLUSIONS 2
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The NRC' staff has reviewed the nuclear systems subtask report and finds no i
new issues with potential safety significance that should be' addressed.
The staff l notes that appropriate technical specifications ensure that problems q
invol' ing reactor operating flexibility and plant capacity'are not alleviated
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atithe-expense of safe operating limits; such technical specifications are in place on-operating reactors and any changesiin reload fuel desi been< identified as a recomme,nded action to avoid power derating,gn, which has:
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NRC review where required by 10 CFR 50.59 for impact on safety. are subject to
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- 5. 2 Subtask'B: = Report on Fuel H
INTRODUCTION I
l The subtask report on fuel. deals primarily with the design and performance limitations of the fuel'and'related core components in the context of their im-l
-pact on the reliability and availability of BWRs.
3 teraction (PCI) of the GE 7x7 fuel was the predominant' fuel problem at the-Be a
time of GE's 1975 study, fuel preconoitioning operating recommendations'and L
f design' changes needed to resolve the PCI problem received most of the-attention.
'There were also concerns that regulatory requirements-based on the ALARA prin ciple'could increase the obstacles to design improvement.and changes through -(
I more, comprehensive and conservative fuel design'models for transient analysis, cations enforcing PCI operating recommendations.more extensive ~ pro NUREG-1285 34 Ie
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SUMARYOF1SSUES$ 'h'
'i The; principal'ihsues in this; subtask report are addressed in the. Reed' Report.1
.The' safety significance.and current status of the following Subtask 8 issues l
'are discussed in Section;3,of this report in the listed subsection.
Subsection-I Issue d
Impact on Fuel Integrity of Reduced' 3.7 Moderator Temperature.due to Equipment x
Failure
' Impact of~ Core Design and. Licensing 3.15 Criteria on BWR-Capacity T
. Possibility:of_ Control Rod Binding due to 3.18' Fuel' Channel Creep 1
1 Seismic Capabilities of 8 x 8 Fuel Spacer 3.21 Invits detailed review of.this subtask report, the staff' identified two addi-~
tional issues that~ warrant attention.
A discussion of these issues, and their L
' safety significance, follows.
End of Life Failure Modes L
Issue:.-Fuel performance data at the time of GE's 1975 study was limited L
to 15 to 20 GWD/T exposure.
There was concern that after resolution of L.
.the PCI problem, failures would occur from exposure-related problems-such
-as 4
fuel' swelling due to fission products' contained in the fuel failure-or distortion of cladding due to. fission, gas pressure fg,
. thermal' fatigue of cladding
. failure of cladding due.to corrosion L
failure of-cladding due te fretting and wear by spacers b
weld area penetration-f Status:
Analytical models_ for design prediction of extended burnup per-iL
'formance have been developed and approved by the NRC staff. BWR fuel has q=
been approved for operation to extended-burnup of 40 GWD/T batch average exposure.- ' Operating experience with BWR fuel in excess of 30 GWD/T has not' revealed any significant performance problems with extended burnup.
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fuel.
-Incipient Cracks a
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- Issue:
Unfailed fuel of moderate exposure may contain multiple incipient cracks, which makes the fuel susceptible to failure under unusual stress.
Safety Significance: This could ca'use under-prediction of core damage and radiological consequences associated with transtants and accidents.
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- Statuu:, Operating experience has not shown' any problems associated with p Q* "
G thu "ailure mechaniset Conservative fuel failure criteria woul'd bound M
such failures if they did occur; for example, in the licensing basis safety -
M, i
analysis' _ any fuel driven to a critical heat flux power level is assumed to y
. fail (a c,onservative assumption).-
h,
- CONCLUSIONS-w The NRC staff has reviewed the fuel subtask report and finds no new issues With potential-safety significance that should be addressed.
The predominant' fuel-problem (PCI) at the time of GE's study has been substantially resolved,'and:
there are no:new problems associated with currently. approved fuel designs or with operation at extended burnup.
5.3 Subtask C:
Report on Electrical. Control and Instrementation Systems' INTRODUCTION This subtask report addressed the design process for the electrical,' control, and instrumentation systems to assess the adequacy of design methods'and ap-proaches to produce the required product performance, quality, and availability.
In addition, design uncertainties were identified and corrective actions-recommended.
SUMMARY
-0F ISSUE _5
.The NRC staff review of this subtask report addressed the specific areas' dis-cussed below.
BWR Dynamic Control System -- Dynamic Control and Load Followino Capability o
issue: The subtask report recommends that GE perform an overall' systems evaluation of-the' technical feasibility of, and the economic justification for, modifying the BWR dynamic control. system to provide increased capa-bility for normal electrical grid frequency control duty and for coping with network disturbances (such as might lead to isolated grid operation).
It also recommends that GE evaluate a joint internal effort in this regard.
Status:
Dynamic control with load-following capability is not generally approved for-BWR plants, but the NRC will review applications for this capability on a case-by-case basis. This issue did not raise any new safety concerns.
BWR Dynamic Control System -- Pressure Control System Issue: The subtask report recommends that GE always have on hand, in San Jose, one. set of qualified pressure control system hardware..so that if:
problems arise overseas, there is a quick and effective way to test and evaluate solutions.
In addition, the report recommends that the responsi-bility for at least the electrical components of the pressure control sys-tem be transferred to GE's control and instrumentation group.
Safety Sionificance:
This issue did not raise any new safety concerns.
NUREG-1285 36 l
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- BWR Dynamic Control: System 4-Automatic Load-Following-System' Issue: Th'e' subtask report recommends that' GE's nuclear engineering. group i
become thoroughly: acquainted with.the~ advantages and disedvantages of-a various electronic variable-speed pump drives for recirculation flow to determine ~if they.might serve as a backup for the flow control valve and t
to ensure themselves thatt the valve system is really warranted in view of-potentia 1' availability advantages of the variable-speed systems.1 In addi tion, the: report; recommends ~ that GE consider, and have designs for, alter-:
a natives to the non-linear 3-mode controller.
1 Status:: This issue did not raise any new safety concerns.-
's BWR Opamic Control System -- Feedwater Control System i
Issue:
The su6 task report does not provide any recommendations concernin this issue.
The-NRC staff has recognized that there are operational pro g--
blems associated with the feedwater control system.
All of these problems =
' fall into;the operational-category (not safety related).
All BWRs will.
include a;feedwater trip to limit vessel-high-level transients as required for the resolution of NRC's Unresolved Safety Issue A-47.
Other.initia-tives.in:important-to-safety balance-of-plant systems such as feedwater systemsLare being considered by the NRC staff.
Safety-Sionificance:- This issue did:not raise any new safety concerns.
BWR Dynamic Control System -- Relief Valve Auomented Bypass-(REVAB)
Issue: The subtask report recommends that GE review the ability o'f REVAB to meet its design objectives and consider modifying;the REVAB operational objectives, ir. light of potential impacts on plant operational availability.
In addition, the report suggests that GE review' alternative means for providing the: capability to accept loss of electrical-load without reactor scram, and' compare them with REVAB (on. technical and economic bases) to form the basis for GE's future approach in this area.
Status:
REVAB has not been installed on any GE BWR in the United States.
This issue did'not raise any new. safety concerns.
Control Rod Drive System Issues:
The subtask report recommends that GE (1) continue its program for fast-scram development. ensuring that it maintains.the required priority, program direction, and resource level needed to make available well-tested drives for initial opera-tion of first BWR-6.
GE should also ensure that adequate develop-mental test facilities are available for testing of prototype drives with blades, under pressure, temperature, clearance, and water quality conditions to be encountered in operation.
NUREG-1285 37 o
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, (2)! initiate a program in parallel with the' present evaluation / redesign-Iy o
c-tof the' control rod drive.
Specifically, GE should evaluate the:
- c potential for a " Vernier motion",added to the planned hydraulic 5 fast-scram drive.
4 V,
Status:' This' issue did not raise any new safety co'ncerns, The design of s
the rod drive' system for the BWR-6 has been reviewed.and approved by the
'l NRC, and the timing of the rod insertion for a scram has been taken anto account in the Final SaTety Analysis' Report for BWR-6s,' and is periodically.
+ '
- verified through surveillance tests.
~
Reactor Safety System -- Setooint Drift
+
Issues: The subtask report recommends that GEJ (1)- continue to give the required priority to this problem and its=
&(,
corrective program to ensure that GE's schedule for issuance of-Engineering Change Authorizations is met U
(2) take the initiative with its customers, and with NRC, to ensure that the required changes are, implemented on a timely basis-Status:
Setpoint drift is bein
,w a setpoint' methodology program =g reviewed by the NRC staff..GE established 1n the early 1980s.and issued NEDC-31336,
" General Electric Instrument Setpoint Methodology," which seeks to confirm -
' c" the adequacy of protection system setpoints, including. allowances'f,or-drift.
NRC is reviewing HEDC-31336.
This issue did not raise any new' safety concerns.
Reactor Safety System -- Solid-State Safety System 3
Issues:
The subtask report recommends that GE (1)' at the proper time in the detail design stage, implement design review-of sepsures taken to. ensure acceptable-' electrical-noise immunity in the system, using some knowledgeable' people from other divisions or outside GE (2) -continue to review the relative reliability of ac solid-state drivers and contactors as output elements to establish expected lifetimes before making a final design commItser.t -
e Status: HRC reviewed and approved the safety. aspects of the solid-state reactor protection s The results of this ystem during the Clinton operating license review..
review are discussed in NUREG-0853, " Safety Evaluation Report Related to the Operation of Clinton, Power Station," dated February, 1982.
The' system is preser.tly operational with no ongoing safety concerns.
This issue did not raise ~any new safety concerns.
p-Neutron Monitoring System
+
Issues:
The subtask report recommends that GE (1) defer to its Task Force 6 for recommendations on the incore sensors n
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' % g% j(2)'sreviewtheTravelingIncore-Probedesigns'toevaluatemoreeffective:
.c g; f m Lsolutions to both the position read-out and guide tube concerns up f~',
Status:. This issue did notl raise any new safety concerns.
z 6;f "$.
-Other Instrumentation Systems
),',
Issue:
The report did not provide any recommendations'; it stated that p,,y y specific problems that have occurred seem to be adequately resolved.
t Status:
This issue did not raise any new safety concerns..
wo l
Power Generation Control Complex (PGCC) l
.Issuei The report did not make any recommendations: regarding the PGCC. -
l Status:
The NRC ' staff ha's re' viewed and approved the PGCC during several
~p operating license case reviews (e.g., Susquehanna, Nine Mile Point 2, LaSalle).
This issue did not raise any new safety. concerns.
NUCLENET-Complex n[
Issues:
The report recommends that GE (1) complete-two technical design reviews on display control system:(DCS).
y in 3rd quarter of 1975 and 1st quarter 1976, utilizing some technical experts from outside the nuclear engineering department.
In the future, these' reviews should be done routinely using such outside experts.-
(2). confirm that its staff is capable of maintaining the first NUCLENET
. hardware system.-
.(3) Emake maximum use of interactive computer graphics for the printed circuit-board work.
(4) obtain early data on the reliability of the 4400 computer.
(5)T explore the opp rtunities *.o use Honeywell-PCD standard software as
- a basis for DCS system.
(6)- = review the plans 'or field maintenance of NUCLENET systems to ensure-that someone is doing the test and diagnostic programming and proce-dures York necessary to keep the equipment operating in the field.
Status:
NRC reviewed the safety aspects of NUCLENET during the Clinton operating license review.- The results of this review are discussed in
-NUREG-0853, " Safety. Evaluation Report Related to the Operation of Clinton Power Station," dated February 1982.
The system is presently operational, with no ongoing safety concerns.
This issue did not raise any new safety concerns.
.NUREG-1285 39 i
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Plant Auxiliary Power Systemsi 6
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-!ssues:-'The report _ recommends that GE g
(1)! give its customers increased application engineering assistance to' J
' emphasize the need for greater main switchyprd redundancy.to improves
(*(
- plant; availability.
(2) 'sp~ ~ecify the redundancy and other specia1' requirements of power sup-plies provided by'the customer for non-safety related GE systems-o affecting's plant s availability.
These specifications should 4
include electrical,- pneumatic, and hydraulic supplies, at all power.
7-levels.=
'(3)c centralize the responsibility for power supplies for all GE systems-f ~
to_ enable an effective approach to power supply / plant unavailability problems.: In addition to documenting and coordinating all power-
- t,
supply-requirements for availability-related systems, an important part of this effort should be convincing the customer =of the; benefits of meeting these requirements.
e g
Status:- Durin applications.(g the licensing review of recent BWR operating license e.g.. River Band, Perry, Nine Mile Point 2), the NRC staff has been' unable to-find consistency in a utility's characterization of the Class IE/non-Class 1E beundaries associated with the' reactor protectionsystem(RPS)powersupplies.
In fact, in some cases.,an in '
i dividual utility has been. confused ~as to the ' location (s) of_ this boundary.
m
- This has led to various separation, physical identification, seismic, and Class 1E/non-Class 1E interface concerns regarding RPS bus A and B.
The' staff believes that if-the third recommendation had been followed for.the RPS power supplies',Lthe confusion'regarding the concerns addressed above would'have been allev'iated.
The staff is reviewing this issue to deter-mine:if 1,t should be considered further, possibly-as a generic issue, g
C&I Availability / Reliability / Maintainability Program Issues:
The report recommends that.GE (1) show a greater concern for and preoccupation with the safety-aspects of nuclear design ~In non-nuclear projects, the safety aspects are easier to address and, therefore, require less utilization of:
resources!and regulatory involvement.-
(2) develop its nuclear projects to the same order of operational reli-ability that customers for non-nuclear projects (NASA, 000, etc.)_
demand.
-(3) encourage greater reliability efforts.
Innon-nuclearprojects,the customer (NASA, 000, etc.) demands, funds, and monitors a reliability program.
In the nuclear industry, NRC provides a reliability stan-dard for protection systems but does not fund the effort.
GE's utility customers are not known to either require or fund reliability efforts.
i NUREG-1285 40 a
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.(4); have:its nuclear engineering department reliability and maintainL
- ability plan objectively reviewed by knowledgeable GE personnel Outside the= department.7 s
A
'(5)L strengthen its nuclear engineering department problem / failure report--
ing system by consolidating the current multiple systems into.a ~
single, comprehensive system with closed loop features to ensure.
6 e' accountabil_ity and satisfactory dispositions.
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-(6)? initiate-education and training courses in availability / reliability 1 a
v.
maintainability engineering so that there is'a more consistent' and uniform approach to these' disciplines in the design' engineering i
community.-
q Status: The NRC staff believes that.the current industry maintenance-program and technical specification. surveillance requirements provide-adequate assurance that safety systems will be available-when required.-
1 There is an ongoing program within the Institute of-Electrical and Elec-l o
tronics Engineers (IEEE) to provide enhanced maintenance guidelines for-
~
-many tvpes of. components.- In' addition, several vendors have submitted techni ale specification-improvement programs to the-NRC.
This_ issue did c
not raise any new safety concerns, t
-C&I Component and Sy' stem Qualification Issue:
The. subtask report recommends that GE's standards and qualifica-3 tion engineering department be given additional-manpower and the responsi-bility for reviewing and approving the qualification of all systems and 1
~
-components for which C&I has responsibility.
s i
-Status:. The NRC has-stringent component and. system qualification standards.
This issue did not' raise any new' safety concerns.
Systems Resp'onsibility
+
4 Issues: LThe report recommends that GE (1) focus the. responsibility and authority for total BWR system design d
specification and control as the full-time responsibility of a senior-i technical manager and a small group of highly qualified system engineers, j
(2) establish the required management and operational policies and proce-dures needed to ensure that this group receives the required support.
from GE's design, manufacturing, marketing, and projects organizations, q
Status: This issue did not raise any new safety. concerns.
1 CONCLUSIONS As a result of its review, the staff concludes that, with the possible excep-tion of the plant auxiliary power systems issue, no new safety issues are addressed in this subtask report.
The issues addressed either involve (1) concerns that have been resolved elsewhere or (2) concerns that do not
-NUREG-1285 41
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aspects of BWR. safety systems.;
and availability for any safety!
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. ; 5.4 Subtask D:. Report on Michanical' Systems and Ecutoment-
'i
' _ INTRODUCTION' 1
the reliability of major mechanical components in th
?
supply system and the impact on projected plant availability.
x l111' containment issues are also addressed ~ Flow-induc d Mark I II..and ioccurring in-reactors;that had been operating at the time of the stud j
e
' addressed, as are the corrective actions taken in response:toJidentified problems _with mechanical systems and equipment and the design qualific i
adequacy of-BWR-6 components that have no operating history in reac:
This report includes an extensive review of nuclea problems ~and design changes on BWR-6 availability, g
y;
SUMMARY
- OF-ISSUE 5 N,
The safety-significance and current status of Reed Report issues rel q
the containment (including main steam isolation valves), mechanical equi)d i.
to J
failures due to flow-induced vibration', and problems with the TIP system a
d p
discussed in Sections 3,and 4 of this NRC evaluation report.
other issues in this subtask, report that are of potential safety significance A discussion'of-6 fcilows.
,N Crosby Safety Relief Valves (SRVs)
Issue:
Crosby direct spring-loaded SRVs'were to be used on BWR-5~and. that were operating at the time of the 1975 study. system It was expected that the Crosby valves would be more reliable because they do not employ a valve sysfem that had caused actuation problems with' the other valves, pilo
,m i
Safety Sionificance:
The SRVs are required to protect the integrity of i
h the: reactor coolant pressure boundary and to limit the severity cf over-L' pressure, transients.
tion setpoint accuracy and ressating without leakage.The prim i
is required between refueling outages, it contributes to. unavailability If SRV-maintenance g
4 of the reactor.
L i
Status:
Testing and limited operational experience have not revealed an
,; a significant o BWR service. perational reliability problems with the Crosby valves for y _
E Flow Control Valve (FCV)
Issue:
1
~
BWR-5 and -6 systems use FCVs in conjunction with a constant sp pump to control recirculation flow.
durability and reliability of the valves.for this applicat, io NUREG-1285 4i!
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I
~ Status: These valves are now' performing satisfactorily >in operating BWRs.
P CONCLUS]0N s'
l The NRC staff has reviewed the mechanical systems 'and equipment' subtask report :
1 and finds no new issues with potential safety significance that'should be -
addressed.
y 5.5 Subtask E:
Report on Materials, Processes, and Chemistry f
[
INTRODUCTION This subtask report' addressed the materials, processes,'and chemical technology necessary to achieve reliability and quality in BWR systems. The report assessed the.effect of materials behavior, processing, and chemistry on plant i.
reliabilitv safety, performance, and lifetime; evaluated the adequacy of material sv,.ection, procurement, application, and cost; and identified critical p
material and chemical: areas for improvement or additione.1 development.
4 7
j
SUMMARY
OF ISSUES l
The report cddressed among other things the. issue of stress corrosion cracking,
- which is discussed in detail in Sections 3 and 4 of this report. 'In addition, it addressed the areas that are discussed below.
Some of these subjects are,
also discussed.less completely in various places in Section 3.
Radicactive' Contamination i
Issue: ' Concern was expressed-that radioactive contamination of piping and otherLcomponents.would build up to the point where radiation exposure-to' plant maintenance and operations personnel would become excessive.-
l This would' require additional manpower and increased costs.
The report l
recommended that more effort should be expended on understanding the L
basic mechanisms of radioactivity transport and buildup, with the aim of
~
L making modifications to reduce the problem.
Safety Sionificance:
This-issue is'related to ALARA, and is a general, industry-wide problem. Although it is not a reactor safety issue, a greatideal of effort has been expended on it.
It should be noted that GE has. developed a prceedure designed to reduce buildup of radioactive con-tamination in piping and surfaces containing radioactive contamination.
There also have been other major industry initiatives in developing and using decontamination processes, with generally good results.
Status: This issue did not raise any new safety conerns.
Reactor Pressure Vessel (RPV) -- Probability of Failure Issue: The report estimated the probability of a sudden disruptive TaiTure of the RPV to be less than 1 x 10 e per reactor year.
This estimate applied to all presently. designed BWRs.
NUREG-1285 43
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the Advisory Committee on Reactor Safeguards (ACRS) a and'
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Thus this issue'did not raise any new safety con 1
, as delineated _in the y,
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. Reactor Pressure Vessel -- LOCA Inteority
)
Issue A detailed' anal
')
M a:st made in 1968;'ysis'of RPV inte rity in BWRs under LOCA cond
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h
'Much more recent~ reviews by the NRC'and ACRS ha M
- conclusions.-
l
]j Status:
of repressurization in a BWR would preclude lo sA, <
I failure.: ' The issue of. pressurized thermal shock in BWRs
,L i
Safety: Licensing Board hearings on the Lime o' 1 i
R issue did not raise any new safety concerns.
A,
, this:
U Reactor Pressure-Vessel -- ATWS Presscres issue:. Calculations of Fiansient without scram (peak pressures under postulated anticipated-y ATWS) conditions have been made within-the past o@y year for various BWRs.
Peak pressure in the.1600 to-1650'psig range have-been calculated for certain BWR-3 plants and considerably lowe for other BWRs.
These pressures are well.within the capacity of.the w
vessel.
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_ Status:
pressure expected during an ATWS event in a 1350 psig, even less than GE assumed in the Reed Report.
N did not raise any new-safety concerns.
.Thus this issue concern is provided in Section 3.)
(Additional discussion on this
-3
_ Reactor Pressure Vessel -- Faticue Crackino
?
e
-Issue:
GE's stuoies provide strong support for the-idea that fatigue cr H growth in vessel steel under BWR environment condition 3
have~an adverse impact on RPV integrity.
within specifications. stress corrosion cracking would not occur in Sta'tus: - Fatigue cracking caused by anticipated transients 1
under ASME Code rules, is very unlikely, even with the known delet 4
effect of BWR coolant-on fatigue strength.
Recent studies also provide assurance that when RPV steel is properly heat treated and stress it is not subject-to stress corrosion cracking at stress levels fou 0
reactor vessels.
The stringent controls on welding and post weld heat-treatments imposed during the manufacture of reactor pressure ve provide assurance that the material will be in a resistant condition high residual stresses will not be present.
, and Thus, this issue did not raise-any new safety concerns.
l L
NUREG-1285 44 4
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eteactor Pressure Vessel -- Nozzle Crackino-O ssue:
Cracks had been observed in' the cladding around feedwater nozzles 11 stone,and Dresden 2, but the cracks were small enough'to be readily'
'A
- a removed.
Ultrasonic-indications of possible cracks at Pilgrim were being-monitored on a continuing basis.
In the BWR-6, the cladding was. eliminated y,
around all nozzles.-
Stat'us:
In 1975, cracks were found in feedwater nozzles of several more' ServIceda'formalinspection.andrepairprogramwasinitiated.
Information Letter No. 207, addressing feedwater nozzle cracking, on November 19, 1976. -All cracking events and repair operations were reviewed and approved by the NRC.
The NRC initiated Generic Technical Activity A-10 to' address this issue'.
In July 1977, the NRC published NUREG 0312. " Interim Technical Report on BWR Feedwater and Control Rod-Drive Return Line Nozzle Crackin " which described'the problem, probable cause, and recommended actions. g,The cracking in both the feedwater
~
nozzles and the control rod drive return line nozzles was attributed to thermal cycling.
Thermal cycling of the feedwater nozzles was caused by an> ineffective thermal: sleeve. GE performed extensive testing and analysis, which resulted in recommended changes in design of the spargers and thermal sleeve. -This was documented in GE's report NEDE 21821-A,.
issued in February 1980. The NRC resol + ion of this issue was documented t
in NUREG-0619, which was issued in November 1980, and was impismented by NRC's Generic Letter 81-11.' This-problem is considered to be resolved.
Reactor Pressure Vessel - ~ Inspection Access Issue:
The BWR-6 was designed to accommodate currently specified and reasonably antic'ipated' future RPV inspection requirements.
- However, inspection of RPVs in older plants, if required, can be performed to only a limited extent with currently available equipment and methods.
Status:- While access to the RPV is provided for examination equipment' in the BWR-6', the equipment.itself had not been fully developed at the 1
time this subtask report was-written.- Further, the ASME Code-specified inspections of ligament areas between control rod penetrations in the bottom head were not then possible in any BWR.
Where such inspections.
-are not practical, NRC may grant relief from the Code requirements.
Regarding the inspectability of the shell portion of the reactor vessel, including the radiation-affected belt line region,- some BWR-5s provided access for inspection. -Preservice examinations of this area have been performed at plants built fairly recently; therefore the equipment for examination from the outside has proven to be practical.
For older BWRs the NRC has granted relief from examination of the major shell welds, because the biological shield is so'close to the vessel that no examination equipment can fit in the insulated area.
BWR vessels cannot readily be inspected from the inside (as PWRs are) because such internal structures as jet pumps are in the way, and the internals are not designed to be completely removed.
GE has a program with an overseas utility to develop equipment and methods to remotely inspect a significant portion of beltline welds by ultrasonics NUREG-1285 45 l
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from the'inside'of: the vessel. J The NRC staff expects that such methods d
will:soon^be' developed.for general use.O However,'the. staff.does not be<
i*
lieve that the acoustic emission inspection techniques' mentioned inithe-k.
, subtask report have been sufficiently developed to be considered a realis-tic and practical. approach. - Nonetheless, this' issue did not raise any new '
t
(
7
. safety concerns.
(
Reactor Pressure vessel -- Embrittlement,
1
- (
i Issue:' The oldest SWR plants (e.g.
Dresden 1. Humboldt' Bay and BigL i
E Point).did not have Jet pumps and have the pressure vess,el closer to the core than is the case with later reactors.
This..has,resulted in.
A higher radiation levels and the potential for a higher degree _of rad'istion ombrittlement than will be encountered in subsequent: reactors. - No operat-i ing problems'are foreseen, but thermal annealing may be desirable at a:,
J later date to ensure that these plants can meet hydrostatic test l
requirements.-
q L
Status:- :Dresden 1 and Humboldt Bay are of no further concern because they have been decommissioned; Big Rock Point does show a considerable.
- [
~
radiation-induced increase in RT Nonetheless, the NRC staff has had:
no'indicationthatBigRockPoinba.s any difficulty in heating up to the 1
required temperature for leak and~ hydrostatiettests. This is partially.
3 becausenthe vessel was designed to Section I.of the ASME Code, so stress levels are very low.
0ther later BWRs are starting;to show-the effects.
'j of irradiation of.the vessel on testing temperaturesi Some licensees have considered the use of-external heat sources to help achieve =the required temperatures.,~However the subtask report is correct in stating that irradiation of.the vessel w,ill not limit operation; thus'this issue
.did noT. raise any new safety concerns.
Matefials Information System and Control
' Issue:..The subtask report discussed the need for GE to establish a l
stronger materials engineering o.rganization with~ better laboratory facilities; Status:
This issue did not raise any new safety concerns.
. 1 Level of Materials Effort
' e'
. Issue:
The issues discussed above addressed specific needs for extra
.itTTGt on' stress corrosion cracking and radioactive contamination by Coso, Other: materials areas exist where continuing, although less severe, prob-less should receive more attention.
Components involved include reactor; pressure vessels, control rods and control rod drives, reactor core inter-nals,: steam separators and dryers, pumps, isolation and safety relief vales, condensers, heat exchangers, electrical insulation, and~ protective coe ings and paints. While active work is in progress in most of the aru s and no significant deficiencies have been identified,-the subtask report; indicated that GE should expend additional effort to meet the high-availability / capability goals on which its strategy is based.
NUREG-1285 46 g
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g4 Status:f 5everal of these: items have been covered in detail elsewhere-(e.g... control rod materials and failure mechanisms, corrosion, Zircaloy channel materials, quality of vendor-supplied components, and radiation.
- damage studies).
In regard to the development of improved gasket, seal, and packing materials although-fewer and smaller leaks would enhance
' plant availability, this is not considered to be a safety-issue; leakage limits already are imposed by a plant's technical specifications.-
In 4
sum, none'of these issues discussed in the subtesk report raised any new safety concerns.
o CONCt.USIONS On the basis of its review, the NRC staff did not find that this subtask report raised any new issues;with potential safety significance.
5.6 Subtask F:
Report on Production, Procurement. and Construction INTRODUCTION This subtask report addresses ' critical colhponents manufactured by GE, componentsLprocured from outside vendors, and the field erection of the
' nuclear steam supply systwm.
l
SUMMARY
OF ISSUES-l:
' ThisL subtask report on production, procurement, and construction identified
-l concerns regarding fuel rods and vendor-supplied components. A discussion of these issues'follows.
- i Fuel Rods Issue: The report identified the following concerns regarding fuel
- i rods:
J (1).GE'should manufacture one standard fuel rod and one standard fuel pellet and compensate for needed variations by using:different 1
enrichments and rod arrangements.
A second rod site may be needed to reduce fuel failure.(increased wall' thickness and reduced pellet diameter) at the highly stressed corner position.
(2) In light of technic'ai problems with fuel rod leaks, GE should review its decision to reduce the_BWR-6 fuel pellet diameter by 0.006 inch and: reduce the fuel rod wall thickness by 0.002 inch.
(3) GE should improve the quality of the zirconium tubing it produces for fuel rods. Although the tubing is acceptable, it is of lower quality than that produced by Sandvik.
Areas to be improved'
-include roundness (it is not consistently round), surface flaws, 4
and inspection equipment.
Status: Appendix K to 10 CFR 50 contains the NRC requirements for fuel rod oehavior during a loss-of-coolan,t accident. A plant's technical specifications establish the limits on the release of fission products from fuel rods as a result of normal operations and transients.
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orated by detailed analysis. and-testing.- This. issue does not raise any new safety concerns.
Vendor-Supplied Components Issue:~ The report identified the following concerns regarding vendor-supplied components:
,a i
(1)L. Vendor-supp11M components are a cause of plant outages.
Specific-areas that must be improved-include the qualification plans and:
commitments 'of qualificationLfacilities, management commitment for.
establishing an: integrated' reliability program, valve testing,~and reliability, analysis in.the design process.
In addition, the report:
suggests eliminatinc vendors who do not provide adequate engineering.
. support and performing studies'of--sufficient depth to support _the quality needed for the nuclear industry.
(2) There was;a high probability that a qualified flow control valve for.
the recirculation system would not be available.for a 1977 startup of.
.BWR-5 plants.
(3) GE should consider manufacturing some components tupplied by.
e vendors.
(4) For the'PGEE/NUCLENET System, GE should eliminate onsite changes
-by-completing fabrisation of the electrical and control, system in 1
the factory rather than on the site.
Status:
Appendix B to 10 CFR 50 addresses the QA criteria for.the~ design.
1 and manufacture of safety-related components.
It also provides the basic' requirements for improved reliability of performance by implementation of the criteria on design control and corrective action.
In additionf the NRC staff conducts an extensive inspection program which reviews the utility's activities and;those of its principal contractors and vendors to determine conformance with NRC requirements and regulations -
Lincluding those cited above.
This issue did not raise any new safety concerns.
CONCLUSIONS On the basis of its review, the NRC staff finds that this subtask report did not-raise any new safety concerns.
5.7 Subtask G:
Report on Quality control System Overview INTRODUCTION j
This subtask report addresses the adequacy of the quality control system utilized by GE for the des'ign, manufacture, and operation of nuclear steam supply systems for BWRs.
It compares this system with the quality control systems adopted by five other GE organizational components.
NUREG-1285 48
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, h,[$UMARY OF ISSUES -
J e
x The subtask report recommended that GE estabitsh a " reliability" organization i to analyze failure and repair data an e favailability goals in terms of design d it discussed a need to establish plantL significant parameters.- It also stated m
that'the resolution of major problems experience on already-constructed plants ~
^ indicated a need for improved designs..in equipment, materials, processes, and a system control.- The report included'a listing of QA audit findings that showed "4
that calibration practices were not formally. documented or controlled, design i
e L "' '
f reviews;and documentation were not-in conformance with established requirements,
' hardware documentation was sometimes not clear, engineers were not familiar with
- ~
>, manuals, and, in some instances basic to ensuring design -integrity, approved
,A engineering practices and procedures had not been followed.
i,,.
- All of these issues are covered by existing NRC requirements and regulations.
D
'Specifically. these requirements and regulations include Appendix B to 10 CFR 50; which delineates the QA criteria-for the design, construction, and
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- operation of nuclear power plants; 10 CFR 21, which requires the immediate reporting of manufacturing defects;=10 CFR 50.55(e), which requires the report-a ing of deficiencies. arising during construction of a nuclear power plant;;and 1
c 10 CFR 50.72, which requires the reporting of certain significant events that n
p occur:during the operation of a nuclear power plant.- In addition, the NRC staff
- conducts an extensive inspection program that reviews a utility's activities g
and those of its contractors and vendors on a sampling basis to determine con-J L
formance with NRC requirements and regulations, including those listed above.
e, 1
It shouW be noted that the Institute for Nuclear Power Operations (INPO) main-4" tains c system for collecting and analyzing failure and repair data.
Access to this-information is available to utilities with nuclear power plants for use in developing availability goals and improved maintenance. programs.
L CONCLUSIONS 4
On the basis of its review, the NRC staff finds that this subtask report did not. raise any new safety concerns.
5.8 Subtask H:
Report on Management /Information Systems
-INTRODUCTION-This subtask report addresses the adequacy of management systems and their implementation to integrate and control BWR operations in the areas of design review,- construction management, startup procedures, project management, and feedback of operating plant information.
SUMMARY
OF ISSUES A discussion of the findings-of this study follows.
o Desion Review L
l-Issue:
Procedures for overall BWR systems design reviews should be.
improved.
9 NUREG-1285 49 a-.,
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Status:: 10 CFR 50,' Appendix B, gives the NRC QA criteria for design '
q construction.-and operation'of, nuclear power plants. -Specific require-)
1 ments: include design control to ensure'that designs'are verified and checked and that. design' reviews are performed.L
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y'n Section 3.16'and:did not raise any new safety concerns.This issue was add 1
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Calculational Models.
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Issue: ' Additional wa i
HTculational models.ys are needed to obtain experimental data ~ to verify reviewed more. thoroughly to ensure, consistency of predictio G
.- Status:: mThisiissue was addressed in" Sections 3.6 and 3.17 safety issues are raised here.
, and no new s
Reliability Imorovement r
tissue: A positive needed to increase,p.high-visibility reliability improvement program is Q
Ef s
lant availability.
~
Status:'.This issue.is=not directly related to plant safety.
'h related areas. NRC regul tions req & e the following:
10CFR21 requires 4
However 'inL 1
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'the immediate reporttig ofcmanufacturing defects; 10 CFR 50.55(e) req tN reporting of daficiencies arising during plant' construction; and-10 CFR 50.72 requires,the reporting of certain significant events that occur during: plant operation.
by NRC regulations, and this issue,did not' raise any new saf In1 addition,-.the study, cited "12 unresolved 238 GESSAR items" that had mentioned in a then-recent (circa 1975) letter:from the NRC Advisory Committ oniReactor Safeguards. 'However, no details of this mention were given.
- the contextLof the report, the concerniis a management and information tran From; problem,;and so has no apparent safety significance.
b a
' CONCLUSIONS
,On the basis.of its review of this subtask report, the NRC staff concludes that NRC requirements and regulations adequately address the safety issues[
mentioned,'and. finds that this report did not raise any new safety. concerns.
a i
5.9 Subtask 1:
Report on Reoulatarv Considerations 9
INTR 00'UCT10N BWR power plants, including loss of availability and ca addressed ways of reducing this impact.
-requirements had added up to 5% in direct equipment coststThe report conclud regulation-induced delays The report estimated that about 15% of GE's and probably'more in engineering time was expen.-ded on licensing matters.
In addition the report-attributert a loss of 2% to 5% in annual electrical generating cap, ability well as'iMeased plant personnel requirements, to the regulatory process,.
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The. study concluded that GE nad contributed to'.the regulatory costs.by failing d
C, to adequately' develop some of the pro required by NRC to validate assump-1
'tions made in'the preliminary design.gress-This resulted in late identification of l
! design problems 1 thus' requiring changes to installations already in' place.
As part of the1 recommendations for reducing the regulatory impacts the study g
-listed potential new regulatory requirements. likely to impact BWR-6 plants.
1
.It also,11sted possible long-term regulatory requirements.- The study recom-l
' mended ways.-that GE could minimize the impacts of these requirements if and
=1 when they come into being.
SUMARY'0F-ISSUES J
A discussion of the issues raised in this report follows.-
h Period of Safety of Unattended Reactor b
i L
Issue:
The study recommended that the GE product safety standards be 4
liio5iTied to-ensure that a. reactor will respond automatically to a reactor
-)
upset or accident to maintain core cocling for at least 30 minutes without y
operator intervention.
The existing standard permitted credit for. operator intervention in 10 minutes.
Safety Sionificance:
The time available for operator-response relates to.
the probability that the required intervention to mitigate the consequences q
of the event will be correctly accomplished.. This is a human factors consideration.
Status:
The NRC staff has some flexibilit y
actions (e.g., suppression pool cooling), y in this area.- For some 1
the normal practice is to accept b
- an assumption of operator action within 10: minutes if it is justified on a L
plant-specific basis.
The NR:: Standard Review Pian NUREG-0800) permits assumed actions within 20 minutes for emergency core (cooling-system l
long-term cooling, and within 15 minutes for' response to boron dilution events.
for anticipated transients without scram, credit'for operator interventi.,within 2 minutes is permitted.
The NRC staff has reviewed
- the BWR-6 to ensure that it conforms to these criteria.
i Means To Identify and Inspect Failed Fuel Issue: The study concluded that the main steam line (MSL) radiation monitor,'which was used for prompt detection ~and shutdown of the reactor for a sudden and major fuel failure, may not be sufficiently sensitive for this purpose because of gross gamma radiation (mainly p s) associated with the steam.
It also concluded that NRC sight require an improved technique for locating failed fuel, possibly more sensitive than the
" sipping" technique that requites opening of the reactor.
The study rec-ommended that GE develop an improved failed fuel sensor, but noted that an NRC requirement for location of failed fuel without opening the reactor was unlikely.
Safety Significance:
The MSL high radiation scram and isolation signals serve to limit radioactivity release in the event of fuel failures.
Safety analyses take credit for the isolation function in the analyses of NUREG-1285
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y' the control rod drop accident.
It is also pertinent k,o ALARA considera-
-tions.~ The procedures and efficiency for location and removal of failed fuel:also are important ALARA considerat.
-Status:: NRC requirements have not' ch :gei. The' BWR. owners' have' indicated an intent to propose e'limination of the.MSL radiation isolation and scram
- )
functions.. This would be justified.by analyses that take credit for gas j
'Of holdup in augmented offgas systems.
Sipping romainsithe most effective
- means'of locating failed fuel, but techniques using this method have M
improved since 1975. Also, fuel failure is much less fre 1975.
This issue did not raise any new safety concerns. quent than in -
4 p
Functional Soecifications for Power and Self-00erated Valves 1
1 Issue:
The study concluded that NRC was likely'to impose spe~cific func-s tional re
-problems quirements-for valves that had a history of frequent operating such as safety; relief valves and main steam line isolation valves. The stud
' specifications. y recommended that GE develop appropriate-Safety Sionificance:
These valves are required to function as assumed in the safety analyses to limit the consequences of various transients and accidents.
Li
. Status:
A TMI Action Plan requirement
-detect leakage of safety-relief valves.provides for acoustic monitors to~
Safety rel.ief and main < steam iso-
. i lation valve 4
led by techni al-specifications.erformance and surveillance requirements are n
.t These are based on functional require-ments and safety analyses provided by the designer of the plant's' nuclear steam supply system and-the specific plant licensee.
sn N-2 Safety Loaic
+
' Issue: -The stue
'i gency cooling sy/ postulated that NRC redundancy requirements for emer-p _
stems.might be expanded to require three, complete systems, each capable of cooling the core in the event of a LOCA.
t With N = number' of.available~ systems,.this is defined as N-2 logic, which permits one sys-tem-to be out of service for maintenance and testing and a second system to fail when needed without loss of the emergency cooling function.
The report recommended that GE study the-need for N-2 safety logic as'is:used in German and Swiss reactor systems.
4 Safety Sionificance:
The degree of redundanc in the emergen core h
cooling systems is related to the system avai ability and pro bility of i
core melt.
Status:
The staff has not identified a need for additional redundancy in-this area, and this, issue did not raise any new safety concern; however, N-2 logic.is an approach being considered by the staff in its study of Unresolved Safety Issue IA-45, " Shutdown Decay Heat Removal Requirements."
Removable Reactor Internals Issue:
The study considered the susceptibility of internal BWR components
-to radiation damage, flow-induced vibration, and other failure mechanisms NUREG-1285 52
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. that'aightf require replacing reactor internals E The difficulty of replac-1 a
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.r and recommended that such a design' be-developed for:later BWR-6 orders and
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for advanced designs.-.
i
- 7 The saf'ty concern of this
- issue isLthe need to y,
Safety'Sionificance:
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ensure that component failures cannot result in unacceptable consequences and that appropriate surveillance procedures and monitorin instrumentation o,
are in place to detect such failures before they degrade p ant operating safety.
Additionally, replacement of failed reactor internals components
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, is'a major ALARA concern.-
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Status:
Many internal. components -- such as feedwatar spargers, -jet pump 4A holddown beams, etc. - have, degraded or failed in'servics and have;been J
B
- replaced.- Occupational. exposure for this type of work has
- 5cin significant I
v i-but occupational exposure to individuals is limited by regulations.
This issue did not raise any new safety concerns.
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Core Catcher 3
[a Issue:
The'st'udy:noted that a core catcher was, a major issue with the P<,
breeder reactor, and recommended that GE study this issu: 9 it could respond to NRC if,a new requirement were developed.
Safety Sisificance:' Prevention.of containment penetration by the molten
[
' core in tie event'"of a'severi accident is.a major safety sissue.
f Status:
La'ter studies have shown that containment melt-through by a molten
? "7 core 1's-less li_kely than previously-assumed.
The staff is continuing its-studies of severe accidents.
These studies include the feasibility and cost / benefit of passive devices such as curbs to contain a molten core, q%
Thus this issue did not raise any new-safety concerns, j
~
CONCLUSIONS:
h L
On'the basis of its review, the NRC staff finds that this subtask report did l
'not-raise any new safety concerns.
Moreover, the report did not present any ideas ~concerning possible new regulatory requirements identified b GE that give cause for the NRC to re-examine its polic in these areas.- B fore imposing'any new requirements,.the NRC routine y considers the impact on power production in relation to the safety benefit to be gained.
F 5.20 Subtask J:
Report on Scope and Standardization 0
~
INTRODUCTION aN This subtask evaluated the GE NED scope of supply and standardization policy in l<
terms of potential impact on overall nuclear plant availability / reliability and operation.
The approach consisted of analysis of plant performance data exist-
'ing at that time to determine the root causes of plant unavailability and the o
options'available to improve the plants by providing a superior quality product
-NUREG-1285 53 H
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and by extension of the boundaries'of NED scope of supply'and services. Find -
ings-of the study were that abot:t.46% of unavailabilit 4 a.
. building, 19% was due to refu'eling and other outages, y was due to the: reactor-Wim '
and only 35% was due-to-balance-of-plant (BOP) issues. With respect to power limitations and avail-i ability in 1974'only, 14.3% of total capacity was lost due to forced outages' and 16:1% was due to scheduled outages.- The reactor scope was identified as-o
.the highest source-of unaveflability;~ contributions by the BOP' area were small.
Power derating as an initial response-to_ alleviate potential equipment failures from new identified problems and to reduce fuel failures from PCI accounted for.
such of-the lost capacity.
f
,+
The study concluded that expansion of the BWR' customer service area with ex-C panded outage service,;better tools,' improved operation, and special programs'
. I for identified problem areas offered the best, potential for improving avail-ability.
The study concluded also that the BWR availability goal. based on-previously established fossil availability was unrealistic because of identi-m fied technical problems and other_ problems not yet. identified.
l.
The standardization effort was expected to be effective only with the BWR-6.-
SUMMARY
OF;_1SSUES E
.The staff examined those itsms lis'ted in the report'as sources of unavailability-or forced outages to determine their safety-significance.and pow fied-in the subtask report were These issues-identi '
PCI operating management recommendations leaky relief valves-leaky MSIV-valves MAPLHGR limitations sensitized stainless steel cracks (major)
. reactivity shortfall feedwater. sparger problems offgas i
-channels operations management All of these issues are addressed elsewhere in.this staff report and, with the-exception of " operations management," have been substantially resolved.
NRC is. continuing to review and-evaluate operations management by individual licenses.-
CONCLUSIONS
. 3
- 0n the basis of its review, the staff finds that this s'M ask report did not raise <any new safety issues.
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