ML20054K465
| ML20054K465 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 06/28/1982 |
| From: | Eddleman W AFFILIATION NOT ASSIGNED |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| ISSUANCES-OL, NUDOCS 8207020226 | |
| Download: ML20054K465 (25) | |
Text
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4 UNITED STATES OF AMERICA June 28, 1982 NUCLEAR REGULATORY COMMISION l
Before the ATOMIC SAFETY AND LICENSING BOARD Chairman James. L. Kelley Judge James. H. Carpenter Judge Glenn O. Bright In the natter of
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Carolina Power & Light Company and
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North Caroliga Eastern Municipal
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Docke ts Nos. 50-400 Power Agency
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and 50-401 0.L.
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Shearon Harris Nuclear Power Plant,
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Units 1 and 2
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6/28/82 amendment to petition to intervene by Wells Eddleman, pro se I, Wells Eddleman, do hereby amend my petition to intervene in this proceeding, polying on and pursuant to and under 10 CFR 2 714 (a)(3), which states "(3) Any person who has filed au petith n for leave to intervene... may amend his (sic) petition for leave to intervene.
A petition may be andnded without prior approval of the presiding officer at any time up to fif teen (15) days prior to the holding of the special prehearing conference cursuant to (Section) 2.751a."
I have timely filed my petition for leave to intervene in this proceeding (Shearon Harris Nuclear Power Plant operating license).
The special prehearing conference is to be held July 13 and 14,1982 (Board order dated June 4,1982 and mo served June 7, 1982).
Fif teen days prior to the holding of that
>o g
special prehearing conference under section 2 751a, is today, June 28, 1982.
I hold that the language " holding" of the conference Om gg means when that special prehearing conference is actually held.
2"
$o$
It is certain this will be no earlier than July 13, 1982, since mo the Board has not changed the date of the conference since its above-l mentioned Order.
Therefore this amendment is timely filed under 10 CFR 2.714(a)(3).
peed 7 age 7
~2-1 Contention #3 is hereby amended by adding the following:
Line 2 of section (F) on page 36 begins with "of" not "if".
The conniection between management problems at NKMRE CP&L's Brunswick and Robinson nuclear olants, and CP&L's ability to safely manage Harris mad includes dhe following: (1) At both Robinson and Bnunswick, CP&L has repeatedly and continuingly failed to meet radiation protection and health physics standards that are anplicable; they have vio14ted their own plant technical specifications by, e.g.
proceeding to do work in radioactive areas without having the plan for such work reviewed and signed off by the plant nuclear safety committee; they have violated NRC regulations by overexposing workers to radiation on numerous oc casions; and these problems m ntinue to occur, despite large NRC fines.
CP&L's management ha s failed to avoid these problems, and failed to correct them, at its other nuclear plants.
There is no convincing evidence the same management at the corporate level will provide better control of radiation exposure and planning for work in radioactive areas at Harris.
Since CP&L has failed to cure these problems in radiation exposure l
at its other nuclear plants, there is reasonable doubt as to the truthfulness of any claims CP&L will do better at Harris.
(footnote an lies to this nage and last nage)
Ay6/15/82,CD&L'satkorneysfinallygotNCEMDA'snameright 1
on one of their filings in dais uroceeding. NCEMPA had changed their name from NCMPA #3 in December, 1981.
They are CP&L's partner in the Harris project. NCEMPA sent notice of its name change to parties in NCUC Dockets E-44 and E-2 sub h36 which moroved NCMPA#3 (later NC-EMPA) buying into the Harris plant among other nower plants. CP&L had lawyers in this hearing before NCUC. Yet, while I've been noting NCEMPA 's name correctly for months, CP&L, their partner, has not.
It would seem they're not paying attention.
This footnote is hereby incornorated into contention #3, above, as an mnendment. It shows how CP&L can and does fail to notice the obvious with respect to nuclear apamationummmd licensing.
It also shows the$r lawyers are human.
But the standard to run nuclear plants must be high, able to detect & avoid most common human failures, or prevent them.
. (2) the management failures noted by Jacobstein and others, including NRC I&E personnel, at Brunswick, show a pattern of massive equipment and operating problems not being fully corrected; repeated failures in design, management, planning, repairs, staffing, radiation protection; and difficulties in getting sufficient quali-fled personnel and using daem effectively.
These problems continued af ter CP&L and NRC staff claimed (1979 remand hearings) they had been corrected.
Since the same CP&L top management organization that staffed, planned and built Brunswick is staffing, planning senior and building Harris, and many of the Harrisjoersonnel have their only nuclear electric experience at Brunswick, Harris is likely to have similar problens, just as a car bought, repaired and of questionable repair ability driven by a qerse#Etemhte mechanic who is a bad driver would be
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more vulnerable to problems that afflicted that mechanic 's old car.
(2a)
Since CP&L claimed to have fixed these problems, and NPC staff agreed, in the 1979 remand hearings, but the problems actually continue, the truthfulness and/or accuracy of the representations made to that remand ASLB are in question.
If past testimony was inaccurate or predicted wrongly that CP&L would improhve (and I think the history of Brunswick since, as detailed in LERs, i
I&E renorts, Jacobstein's investugation, and other recorts yet to be completed, e.g. the CMMP msnagement audit of CP&L, shows that),
so are CP&L's claims and credictions in its FSAR and nanagement j
capability renort in this nroceeding.
CP&L is acting like the Brunswick plant was run by some other comnany with no connection to the Harris plant.
In fact, the same company runs both, and its same character applies to both, particularly the abilities,
attitude toward safety in oractiSe, and knowledge of its top nuclear i
_q.
management.
If statements on the previous record in this case prove to be false, incomolete, or inaccurate (i.e. in the remand hearings of 1979) that puts a special need for inquiry into the managenent capability issue in this operating license case.
(3) CP&L's difficulties in ouerating the 3 nuclear units they now have, raise the probability that they will be even less CP&L's resources able to operate a total of 5 nuclear units safely.
Irwitaminut evidently are not adequate to even straighten onut the Brunswick olant, ThamRBhiMBanx M much less to do that and safely run Harris & Robinson.
(4) The Robinson plant, which CP&L did not design or build, has operating problems, particularly in radiation overexposure to employees (repeatedly) including contractor employees, and in its dedication to production first, safety second as NRC insoectors have stated.
The Brunswick plant, which CP&L was more heavily involved in the design of, and which CP&L oversaw the building of, and which CP&L ouerates, has even more eroblems:
design flaws, release of radioactive material unnonitored, operation with large percentages of failed fuel (up to 15% or more,as noted
-79 by NRC both in 1978 and in 1981), a host of repair problens, very large number of LERs though CP&L is not shown to be repairs not properly carried out Thfazakamaxa Thazzmdanham y
especially careful to recort everything wrong.
Indeed, CP&L Brunswick has often been very late reporting " events" as at LEPs which require such renorts.
Sone such late renorting continues.
The conntection is that the above facts show a pattern:
nucle ar The more CP&L is involved in a plant (of the 3 they have so far) j the worse a plant it is.
Harris is next in line, and CP&L is even more involved in Harris than in Brunswick.
Worse, CP&L's Harris senior plant site staff, as noted above, nostly have their only nuclear electric experience (if any) at Brunswick, a poorly
-5 managed plant.
(5) The conntection between pay cuts at Harris and poorer nerformance in includes the following: Most peronle nrefer to make more money, rather than less.
CP&L sernior managenent has b een getting widely publicized raises in the 20% per year (and up) range.
Yet, to save money, CP&L gives its Harris trainees being transferred to the Harris site at the same time (spring / summer 1982) a 15% pay c ut.
It is perfectly reasonable to think some of these employees whose pay was cut would resent it, and perhaps perform worse, though they might not admit any such feelings to aCP&L survey or one that identified them by name.
The connection with Brunswick, and with CP&L management capability, is that CP&L gave pay cuts at Brunswick in 1974-5.
n (Thema, the Emptop CP&L management took pay cuts too)
These cuts contributed to CP&L's staffing problems and inability to run the Brunswick nuclear plant with its full defense in depth against nuclear accidents.
The Harris situation shows that CP&L management has noc learned fron that mistake.
Rather, they reneat it in a worse form, by giving pay cuts to Harris employees while taking hefty raises (above inflation significantly) as too nanagers.
L This action shows a disregard of conmon sense and basic hunan feelings of employees also.
Such disregard is not an attribute of good management.
(6) the very large number of licensee event renorts from Brunswick show that plant has many difficulties, from safety-related systens and the ECCS to record-keeping important to safe operation.
CP&L has claimed they " report every little thing" (or words to that effect) in the 1979 remand hearings on Harris
. l CP re managem9nt capability.
But the evidence, e.g. the large number of LERs triggered by NRC innspection, the late filing of LERs, etc, belies this bare assertion by CP&L in that hearing.
As noteds above, these problems continue -- lots of LERS and some late-filed at Brunswick.
And the connection to Harris is that CP&L hasn't been able to straighten out problems at Brunswick (at least not much faster than new ones are foundi, and of ten more slowly), yet many of the Harris senior plant site staff came through the Brunswick " training ground" (or mistake-training ground) and will have to be solving problems at Harris.
The LERS at Brunswick increased during operation as comoared to construction and then stayed up, though with some variability.
(7) CP&L's entire record in running nuclear plants indicates that the management attitude, from the 1960s until the recent period surveyed by Jacobstein (and after), has been one of empha-sizing production over safety; of delaying repairs; of making
,;1egal defenses and excuses instead of making sufficient changes
[
in qquipment, personnel and procedures to ensure safe operation of nuclear plants; and of assuming that they can expand their nuclear operations as f ast as they can raise the money and build only later the plants, and will fix the problems those plants have.
This is shown in the comments of I&E inspectors who've inspected CP&L plants, in their LERS, in their corporate actions and inactions, I
and in their filings d th NRC.
While setting out a facade of I
compliance and cooperation, CP&L has failed to competently design, repair and operate its axisting nuclear power plants.
Their past claims that this pattearn has changed have prohved shaxky and false.
Better ev yet another nuclear p{dggee is required to let them operate
. Contention 61B:
The NUREG number typogranhical error, I think.
A THe NUREO referred to is an EIS for a uranium mill, issued by NRC within the past year.
I can't lay my hands on it now to get the actualy number.
It may be 987. W O M. N A
- O k IM b l}'S NQ gQh tttd CD/ W Y2[ef.5 ha Contention 37(d) is amended to include the following:
"The biased studies and erroneous calculation of food chain and other ecological concentrations are described in NRC translation 520 and other work of the IFEU at ihm Heidelberg, W. Germany, and in the work of icther investigators, e.g. Takeshi Seo, and others cited in Contention 29,e.g. B. Molholt.
Contention 29 is amended by adding to each of parts A,B and C the following:
"That elevated radioiodine levels attend accidents of Class IX and below(some of which may be considered routine operating events) is shown, e.g.
in the work of L. Van Middlesworth, as renorted in Helalth Physics h0:525-527, and elswekwhere, mand in the LAND / LEAF study (John Gofman, chief cons ktultant) of rioutine radiciodine releases fron reactors in and around Wisconsin, which study was published in Methodologies for the Study of Low-Level Radiation."
Contention 65 is amended to addz the following: Daniel International has a repeated record of leaving voidas in containments and base mats of nuclear units; CP&L's surveillance and supervision of Daniel is suspect because of using unqualified inspectors, and because CP&L surveillance and supervision at Brunswick has been deficient in both QA/QC and repairs, as shown in Jacobstein's report (NCUC Docket E-2 sub 428 & elscwhere) and the record of the 1979 CP&L management capability hearings, re supervisors
. not being present much inside the plant, and re CP&L not adequately supervising repair and contractor personnel.
Contention 64 is amended to add the following to section (c):
"By this it is neant that the more irradiated snent fuel that is handled, the nore likely an accident is that involves snent fuel.
Shipping spent fuel to Harris, and then to a Dernanent disposal or reprocessing site if one is found, involves handling the spent fuel twice more unknida -- once to load it and once to unload it at Harris -- than would just keeping it at its point of irradiation (Brunswick or Robinson or other)."
(There nay be another xxxd smendment to contention 6h later in this filing, off this page.)
The following is added to contention 64 after the first sentence:
"FSAR sections 9.1.2 and 9.1 3 merely assert that the spent fuel pool is built in accordance with regulatory guides.
They do not prove this or provide sufficient basis to establish that as a fact.
And the entire FSAR section 9.13 3 (pp 9.1.3-4 to -6) does not provide proof that cooling water will always cover the spent fuel.
E.g.,
if offsite power fails, the fuel cooling numus must be manually connected to the power source.
The redundant buses are stated to involve Unit 4 (for the Unit 1/h pool) and Unit 3 (for the Units 2 and 3 pool) power supplies, which will not exist.
At page 9.13-5 it is stated (no source gi':en, no calculations referenced) that the " slow heatup rate of the fuel pool would allow sufficient time to take any necessary action to provide adequate cooling using the backun provided"... and goes on to say that these " valved and flanged emergency connections" (the backup) are for use during construction in the interzim period in which only one Unit is onerational.
Most of the backup systens for putting water into the scent fuel pools require either manual actuation or nanual connection.
Such many not occur under severe accident m nditions either due to high radiation levels near the requisite connections, or due to tnere being so many alarms annunciated that not all can be kent track of, or because so nany things need doing that there are not enough personnel available to do then.
The number & backgfound required of peoole to actually make manual connections of water waystens is not established in FSAR 9.1 3 3 Nor is it assured that enough people will be onsite at all times to do this."(ref PSAR vol 16)
Contention 6h is also amended to add the following to each of sections (d),(e),(f), (g), (h), and (1): "This information is stated in an article by Dr. Marvin Resnikoff entitled 'Shipoing Casks: Pressure Cooker on Wheels ' published in the Sierra Club Radioactive Waste Campaign publication The Waste Paper, Vol 4 No. 2 at page p 4, Spring 1982.
That article is a " preview" of the oingoing Council on Economic ?riorities study of near tern ootions for handling irradiated fuel.
As of 6/28/82 I understand the comeleted study will be nublished in October 1982."
Contention 25 is hereby amended to add the following:
The suggestion of storing fuel on-site (e.g. at Brunswick and Robinson) for the operating life of a nuclear nlant, and bringing it off in one unit train at tne end, is that of Dr. Marvin Resnikohtff of the Council on Economic Priorities, 84 5th Ave, NYC N'I 10011.
Certainly, the transport of spent fuel is at issue in this croceeding in tha t, if Harris does not receive an onerating license, it cannot receive spent fuel, Thxx so f ar as this nroceeding is concerned, since Applicants have applied herein for a license to
. possess and store Brunswick and Robinson spent fuel at the Harris site.
Thus, the transoort.of irradiated spent fuel to dhe Harris site is an inpact and result of the issuance of a license to operate Harris as the Applicants have requested in this proceeding.
Contentions 2h, 26, 27, and 28, mad 126X, are hereby amended by adding to each of them the same language that above is added to Contention 25 STEAM GEI.'ECATORS:
Contentions 19,112,113 and 114 are each by adding amended taxmid the following:
The AVT (all-volatile) steam
/\\
generator \\, water chenistry treatment xxx7=wtg* (to be used at secondary '
Harris ) was originally rejected by Westinghouse as posing too great a corrosion risk.
And indeed, fx rather ranid corrosion is taking place in Westinghouse plants similar to Harris (e.g.
North Anna 1 and 2, which have Model 51 steam generators --
see NUREG-0886, Table 1) where only AVT treatment has been used.
Note the 2.8% plugged tubes in North Anna 2, operating only since 8/1980.
According to a 1.7.1980 release of the North Anna Environmental Coalition, Westinghouse and VEPCO had provided information to the ACRS and the N. Anna ASLB that such leaks would not hannen at North Anna.
Yet there was suntort plate /
tube corrosion or cracking found for 306 of 888 tubes inspected.
Defects in steam generator tubes increase radioactive releases (by leaks) particularly radioiodines and noble gases.
And they increase worker exposure in testing, cleanup and repairs. (see Table 6 of NUREG-0886).
According to SECY-82-72 of NDC, p.8, N90 is primarily concerned with requiring adequate in-service inspections (ISI) and maintaining xtsmangum primary sys tem integrity, while
, industry has been concerned with developing fixes to oroxatalong 1
steam generator service life and reliability, Up to 25% of so me plants ' annual occupational exnosure (p.3) in recent years has resulted from routine stean generator inspection and main-tenance and as high as 60% for steam generator replacement.
(This is particularly worrysone due to CP&L's continuing record of overex;osure and health physics failures at Robinson, a t'JR that Apolicants have stated is basically sin 11ar to Harris, e.g.
in NCUC Docket No. E-100 sub 40.
CP&L's radiation control failures at Brunswick, and the difficulties of health ehysics there as documented in the Jabobstein report (NCUC Docket E-2 sub 428, etc) and in CP&L and NRC records, likewise indicate this may be an even worse problenn with CP&L plants, such as Harris.
Some manage-ment, same conpany, same de facto policy, of ten same personnel. )
SECY-82-72 says (p.3) that during costulated accident conditions such as main steam line break (MSLB), feedwater line break, or LOCA, the Steam Gene rator tubes are subject to increased pressure differentials and possible pressure waves, e.g. subcooled decompression phenomena,and vibrational loadings, loads e
which Examinerease the potential for failure of degraded S.G. tubes.
l Such failures could exacerbate the accident sequences, provide a leakage oath from the primary to secondary system, X"and several potential leak paths to the environment would then exist."
Steam generator tube failures would also " create a primary to secon-dary leak path which aggravates the s tean binding effect and could lead to ineffective reflooding of the core" in a LOCA.
Even less severe "snall and internediate siz's MSLBiz or LOCAs" whith " tube ruptures leading to or following such events" "could have serious
. consequences.
This is particularly true if fuel damage has occurred as at Three Mile Island. "
Of course, such fuel damage can result from failure to supply adequate cooling water to the core, which in turn could result from (or be more severe due to) steam generator tube f ailures.
At p. 4,SECY-82-72 goes on to state that "An effective solution to S.G. tube degradation problems would require major changes in S.G. mechanical design, thermal-hydraulics, materials selection, fabrication techniques, and dhanges in the secondary system design and operation.
Elimination of S.G.
tube degradation requires a systems approach integrating all of these considerations.
There are no simple corrective actions."
Moreover, (p.5) "the ma jority of the plants under review for operating licenses have S.G.s of similar design to those currently in operation, so that the potential for S.G. tube degradation exists in those plants as well." Harris plant S.G. material is similar to existing plants, see Table 1 of SECY-82-72. It 's vulnerable Moreover, "the consequences of nultiplo tube failures in excess of the design base have not yet been rigorously studied. "
but these are being addressed as part of the TMI action olan, item I.C.1.
The FSAR TMI appendix does not xxxxxx address this item; it aupears to only cover near-term lessons learned, and those not that well.
Much of it is pronises and further review promised.
SECY-82-72, pp5-6 indicate data is now being ceveloped on Astress corrosion cracking and service life of Inconel 600 (the Harris S.G. tube alloy).
That this data doesn't yet exist adequately to provide good modeling of S.G.
tubes made of this Harris alloy (Ianconel-600) is a logical consequence.
- Thus, the life of the Harris S.G. tubes is not established adequately.
. Azuccording to R. Garnsey, Nucl. Energy Vol 18, no.2, April 1979
" Corrosion of PWR steam generators" the condensate colishers can be a major source of corrosive salt to the s team generator tubes.
Further, mild steel for sunport plates "has been shown to be unsatisfactory" as a material.
Japanese work (see " Corrosion Resistance of Inconel-600 alloy under AVT conditions") suggests that a high frequency of consdenser tube leakage, or high levels of dissolved oxygen in water, make the corrosion of nild steel worse, thus increasing denting. (Mitsubishi Technical Bulletin, Nov,1979).
Other than some sumnary statexments in 10.4.6, the FSAR does not establish that these conditions will not occur at Harris. No analysis of operation is provided, just presumed operating numbers, with no renort of tests that establish these numbers in practice, e.g. for dissolved Na, C1, oxygen, and so on.
The FSAR also fails to adecuately consider the electrechemical cell corrosion between the conxdenser tubes (2 different conper alloys, FSAR 10.h) and the Inconel tubes.
The AVT treatment daemicals tend to dissolve out copper, which then can plate out on the condenser tubes or the steam generator tubes.
On the S.G.
tubes, the corrosion cells thus created between cooper and the tube alloy (nickel-base) corrode the S.G. tubes.
In the condenser, such corrosion cells increase leakage, which in turn increases dissolved oxygen and salts which in turn corrode the S.G. tubes, support plates, and other parts of die steam generator in the ways noted above, leading l
to greater orobability of, and greater number of, corroded S.G.
tubes, dented S.G.
tubes, and ultimately, S.G. tube failures.
i (Jesse Riley, Charlotte, NC, brought dhe corrosion cell phenomenon to my attention.)
. What all of the above shows is that the Harris plant S.G. tubes are vulnerable to corrosion, c racking, and leaking which can exacerbate accidents, cause accidents, have the potential radioactive material to release rudimiten to the environment, lead to increased radiation exposure in inspection, nlugging, sleeving and repair /
replacement of S.G. tubes and the steam generators themselves, and lead (as SECY-82-72 notes at,p.1) to additional outage time, which means lower newer output and higher repair costs. (" Approximate-ly 23% of non-refueling outage time has been attributed to steam generator degradation.
The cost of such outages in terms of replacement nower alone is very high.
However, perhaps the greatest financial costs incurred to date are those associated with steam generator replacement."
Secy-82-72 goes on to state tha t the Surry repair / replacement cost about $200 nillion, Turkey Point's similar repair (both 2 units) will be about $6460 million. )
Since the Harris steam generators are vulnerable to such degradation (SECY-82-72 p.5) due to similar materials as existing plants, and due to " shake and break" vibration nhenomena, e.g.
like those at the Krako Yugoslavia reactor with DA 4 steam generators, same as Harris has, such replacement for Harris 1 and 2 must be considered as to likelihood and cost, which will escalate up from the $460 million for Turkey Point.
FSAR 5 4 2.5.3 fails to describe the testing used on tubes re flow-induced vibration, to tell how it is different (if it is) furon testing Westinghouse or others may have done on flow in the D-2andD-13steamgeneratorsggorto startun at McGuire, Summer, Ringhals, Altiraz, etc, to state the conditions of the tests, the type of model used (or was it full scale ?), etc., or in any
. substantive way to demonstrate d e basis, actual tests, actual models, identified tests, or data on which its conclusions are based, when and where and by whom those tests, nodels, etc. were carried out, whether they have ever been checked or duplicated (or refuted) by anyone, what the credentials of the people doing the tests were, the test nethods and materials, andbasis in theory or practice of models and equations used (or references for them) for the claims nade about the steam generator vibrations.
No doubt similar claims are made in the McGuire FSAR.
FSAR 5.h.2.13 consists nostly of clains of afavorable operating experience x and tests with Inconel 600 under AVT.
It does not identify the studies beyond "model boiler tests being concucted by '.iestinghouce", the date to which they have "shown quite favorable results" or what those results were; the laboratory " isothermal tests" (not f urther identified) with high temperature water under " engineering stresses" not further specified, with " production heats" (from the same heats the Harris plant Inconel-600 is from? Else, it's questionable.
Mitsubishi Tech. Bulletin, Nov. 1979 on Inconel-600 shows a good anount of alloying variability in Inconel-600 sanple, including anaxdemmentzmuhuttaxtximagamiatartitmumAdW other incurities nanely aluminum and ti taniun.
Their effect on corrosion, and the effect of variability of the alloy Inconel-600 in terns of varying anounts of its constituent elements as soecified, are not discussed in the FSAR).
In other words, the FSAR on these tests and experience reads like a grant aoplication --
lots of nice things, but nothing to back 'Jhem up.
In view of the vulnerability of Harris steam generators to corrosion and leaks, and the f act that such leaks can cause loss-of-coolant accidents, complicate LOCAs and other accidents, prevent or reduce core cooling, and release radioactive material from the core into the secondary system from which it can escape into the environment, these important facts need to be better established than the FSAR, which reads like a 'destinghouse/CP&L "used nuke" sales pitch.
REACTOR / fast fracture / thermal shock Contentions 46,47,h8,h9,50,41,92,130 and 131 are each amended by the addition of the following:
FSAR 5.3 1.1 does not establish the vanadium levels in the base metal and the welds of the reactor vessels for Harris 1 and 2.
The vanadium content can affect the radiation-induced embrittle-ment of the vessels.
It does not document the tects of the conper os and phamphorus levels stated, particularly for the welds and base metal.
I have seen a statenant that the we1xds on the vessel are in the 0 3% Cu range, which is the range vulnerable to aneutron induced embrittlement according to Dr. Steele of the National Bureau of Standards and other met &1lurgists I have consulted with.
I believe it is from the FSAR but have not located the information yet.
Contentions 47,48 and 49 are hereby amended to add the following:
Applicants' inservice inspection plan has only recently arrived at the LPDR (reactor vessel material inscection plan), about 6/22/82.
even though it is dated $/18/82.
I could not logically have commented on it on 5/14, and neither the LPDR, nor the PDR (which l
I visited June 9 and 11, 1982) had a current enough accession list to show me before 6/22/82 that it existed.
With respect a to its specifications, it is inadequate to detect cracks as described in contentions 47, 48 and 49 because it exempts fron checking by radiography etc. " Specific regions of the weldg and required volumes (which) may not be examined due to part geometries, e.g. core sunnort lugs, flange ID taper, nozzle bores, bottom head instru-mentation tubes), laminar defects, and ultrasonic exanination effects {e.g. near field effects, cladding / parent metal inter-faces).
Ultrasonic examination will be performed on as nuch of the welds and required volumes as practical (as defined by 10 CFR 50.5(fa)(2)(no period, sic) details of the unexanined areas will be provided in the final exanination report." Alternate testing: none."
Since the actual areas not tested have not yet been specified in any greater detail, obviously they could leave out an important area.
Cracks in flanges, bores, and lugs could of course lead to leaks mnd anall LOCAs less than a failed vessel.
But these exceptions even ads drawn are overly broad.
For example, the FSAR states that there is cladding wherever the vessel material is exposed to coolint.
Thus, a large area of the vessel comes under the ultrasonic exception above.
These excentions are so vague that it is difficult to specify all their deficiencies simply because what is and is not to be tested is not stated.
10 CFR 55a(a)(2) simply provides that either the pronosed alternatives provide an acceptable level of quality or safety (no alternatives given, thus they provide none) or that the tests be done except when doing t ien would results in hardship without ann conpensating increase in quality or safety.
CP&L presents in this document no analysis of why each area not examined meets this standard.
. Such an analysis a would be required to enstablish comuliance with 10 CFR 50.55a(a)(2).
But there is none here.
There is no assurance that future examinations will be performed in such a way as to be directly comparable to the data obtained pre-service.
For examnle, will visual testing be practical when the reactor vessel has been irradiated?
Likewise liquid penetrant testing?
Can it be done in irradiated parts and welds (with induced radioactivity and contamination by radioactive material) by humans?
Is this procedure automatable in a radioactive environment? Will it give comparable results as an automated procedure?
Until the answers to these auestions are clearly established as facts, and the comnarability of the preservice and inservice inspections are shown to be real, not just a summary claim in the 5/18/82 letter from Howe of CP&L to Denton of NRC (8205210266 control number), this procedure is not acceptable.
One has only to look at the difficulties of comparing the preservice and 10 year inspections at Oconee (see guardedly worded but revealing letter 8/24/81 from JH Smith, ORNL to W. Hazelten USNRC, at p.2 re sensitivity variation with angle not being established and at p.31st paragraph "there is no good way to compare these data with previous baseline inspection results... This seems to be typical of most reactor sites that I visit. The so-called baseline inspection is usually performed by a different method than the in-service inspection.") to see what problems dhis can raise: lots of data but knowing nothing, particularly of the small cracks Cottrell l
and others believe it is vital to detect.
4 Smith makes another imuortant observation: (p.2) "In summary, I
I
. once cannot determine merely by reading the test procedure whether or not the ultrasonic inspection of the welds to the reaactor vessel will be performed in accordance with Regulatory Guide 1.150."
(Logically, this holds for visual and penetrant and radiograohic examination also).
II smith, an expert, can't make this judgment based on the procedure.
Thus I, a citizen, can only say that his statement shows that CP&L has not established compliance with Regulatory Guide 1.150, nor can the " position" CP&L promises in the 5.18.82 Howe le tter referenced above, sometime in June,1982, provide the information needed to establish compliance with said guide.
Therefore, CP&L must do more to establish that its preservaice and inservice inspections will provide comparable data and comply with 10 CFR 50.55a(a)(2) and Segulatory Guide 1.150 and other aoplicable regulations and codes, e.g.
as referenced in the preservice inseection plan.
This vital baseline inspection, and future inspections, to be of use, must be comparable and actually detect cracks that could cause a fast fracture or other leak.
CP&L has not established that they will comply with applicable dodes, did comoly with them, or that they achieve what is necessary to protect the oublic i
health and safety, ad laid out in this contention.
O ph H-82 2 Or) k tEvojer) Ct YL contention 116 on fire protectionj Sp amendedf F addjkhe following :
The fire hazard analysis of section 9.5A (Appendix) in the FSAR does not address the availability of control and power to the safety equipment.
In general, the FSAR fire section 9.5 is vague, referring to " fire resistive" or " Fire resistant" l
naterials without suecifying them or how long they resist fire in the cable trays, relying on sunface flammability tests not L
. shown to represant actual plant conditions or comr arable conditions, ignoring the possibility of hydrogen in the contatnnent when a fire breaks out (if the igniters don't get to it first, the hydrogen can spread the fired, and the igniters are not lonated so as to prax burn up hydrogen before it can spread).
Ano1her vague statement is that fire barriers are used "kkman w3ere practical" without defining practical or stating the criteria to decide where a fire barrier is or is not oractical (and what type of fire FSAR 9.511.1.1 barrier should be used).
Again, hydrogen could " bridge" around the fire barriers, as could solvent fumes ignited in a a vanor explosion or ranid deflagration.
The " analysis" of Apoendix 9.5A does not denonstrate, as 9.5.1.1.1 claims it will, the adequacy o" other fire protection neasures in all cases.
Rother, it estinates the 3TU of conbustible naterial, smoke generation and renoval rate f aom the area, gives usually a qualitative descrintion of some neasures to mitigate or reduce fire effects, and assumes that the fire will be promptly detected (usually, no analysis of location of detection instruments, etc. ) and the fire brigade will r esnond ranidly and out out the trire, or the automatic ec.uinnent will work.
These assertions are made despite the time it takes to get people into the containnent and to the fire (not well analyzed).
- Further, the " analysis" of what happens if the fire spreads is generally a rationalization that it can't spread nuch, not an analysis.
See, e.g.
" Analysis of Effects of costalated fires".
- Somehow, the effect of a larger than costulated fire doesn't get dealth with in realistic terms.
The whole analysis doesn'tr deal With fires during an accident, when containment would have to be isolated.
Yet it appears such fires ha7 pen (e.g. at TMI,
. the burnt parts of the polar crane.
And I have pointed out the possibility of electrically-started fires and hydrogen burns or exnlo sions, which could ceuse a w1 der-spread fire.
The niant firefighting capability for sinultaneous fires is inadequate, or at least unanalyzed.
This is " Titanic syndrome", i.e. the idea that an event that would overwhelm a "foninroof" systen car.not happen.
3ut, e.g. through the edQiun of hydrogen or of solvent vapors, fires can be spread, and could knock out both parts of a redundant safety systen.
Further, the fire analysis does not censider the effect of spurious electrical signals produced by a fire in cable trays, control equipnent, or the plant comnuter or control room.
Nor are the effects of firefighting on nanual controls analyzed 1
though the ability 'of a person to back into a switch or control and trip or flip it is well known.
This is particularly likely to hanpen in a situation.such as a fire where the unexpected i
nay occur and a person's. attention is likely to be all on the fire and less on where the person is.
If one has to_ back up -suddenly, especially in the control room, m.any sourious signals could l
l be produced this way.
l l
by adding l
Contention 128 is also anended taxxkxim:
The.TMI aupendix to i
the FSAR, p.27, nerely says that Harris has redundant (i.e. 2) hydrogen recombiners.
If. power f rom the " unit h" (nonexistent)'
energency bus and diesel generator would be required, these could not operate in the event of less of offsite power and failure of the Unit 1 diesel generator.
Contention 116 is amended to read "conpuger and/or gentrols for safety equipment" wherever it reads "ICS at presen..
i
. Contention 116 is also amended to read "conputer and/or controls and instrumentation for safety-related equipment and plant shutdown" wherever the plant conputer is referred to as a " computer", i.e.
the oreceding quoted chrase "cor.puter and/or..." is to be substituted for the word " computer" wherever that word referas to the plant cor. muter (I think this is throughout contention 116).
Contention 132 is amended by adding: At TimMI appendix tage 19, CP&L states that reactor vessel level instrumentation designs
" presently available are neither reliable nor unambiguous", and implementation "will be pamtidad forwarded in a future amendment to this FSAR", i.e.
unambi uous core water level readings are not 6
now available.
At page TMI-15 CP&L states that the subcooling indication logic "are not directly testable at power" but the calculated parameter of margin to saturation can be verffled using steam tables.
It is not clear that operators or others will be trained to do this, or have tine to do it.
The incere thermocouoles not being qualified (p. TMIo15, top) means that accurate indication of core conditions may not be available.
Instead, the core condition will be calculated based on core eximit temperatures and the RCS (two cold legs and two hot legs) tenneratures and RTDs.
Thus, actual conditions in the core, e.g. a steam bubble, occurring in a hot spot due to core danage or a failure of a control rod to insert, or a coolant blockage, e.g. from debris lef t in the primary systen during repair and overlooked, night not be shown by this instrumentation.
The in-core part of it is not safety grade and CP&L evidently doesn't propose to qualify it or read in-core pressure and tenperature into their logic.
This means
. the core could have steam volds in it due to uneven tenperature distribution, and the safety-qualified systens night be unable to give an unanbiguous indication of this condition.
Contention 81 is anended by adding:
The " test" of FSAR 13 3.8.1.5 nerely asserts that L test will be conducted and at minimum do testing of communication lines and ascertain availability of transoortation (for whona, not soecified) with mininum disruption of nornal events.
This is not an adeauat e test and CP&L doesn't claim baldly that it is.
Any adeouate test would establish the ability of the plan to go into action at the tire of the test, preferably without warning of the time of the test.
Contention 127 X is anended to state "The resumes of Harris plant site supervisory personnel" instead of X"The resunes of Cs tawba Plant Supervisors".
Contention 41 is anended txxadd by adding "FSAR section 17.2 describes the job descriptions of senior QA personnel and gives an outline of what the program does, but it does not specify the " procedures written" in sections 17.2 3 through i
17.2.17 (not all of these use dae term; most do). The description is a cartoon which says what will be done and ensured, but not haut how it 's Soing to actually be done.
E.g. how are procedures written and anoroved?
How is it verified they do what they say i
i they will?
Does the audit progran involve randon sampling adequate t
to identify errors at a given frequency rate of errors?
The answers to these questions are sunnary descriptions like "An audit planning document is used which identifies... the frequency of the audits... reviewed and uudated periodically. "
It is not
. possible from such statements to establish the adequacy of the QA program, its audits, or any other part of it.
Yet most of section 17 2 of the FSAR consists of such statements.
At best, they are a rough outline of the QA plan, not an adeounte plan nor proof the plan is adequate.
In the light of CP&L's severe QA problems at Brunswick, more paroof by f ar is needed that CP&L QA at Harris is really much better and te ets all anplicable standards or is promptly corrected to meet them so that deficiencies and problems & not recur.
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CE9TIFICATE OF SERVICE I hereby certify that the following have been served with a cony of the attached, by first class US mail, postage nrepaid, (or where indicated by an asterisk, exnress mail; or where indicated by a double asterisk, hand delivery).
day of f921982
[O3 g,
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This, o Wells Eddleman Secretary of the Commission Attn Docketing and Service, 50-400/401 US NRC Washington DC 20555 (3 copies)
Stuart Treby Counsel for NRC Staff Dockets 50-400/401 USURC Washington DC 20555 Judge James Kelley Atomic Safety /401 0.L.
and Licensing Board Dockets 50-400 USURC Washington DC 20555 Judge James Carpenter Atomic Safety and Licensing Board Dockets 50-400/401 0.L.
USNRC Washington DC 20555 Judge Glenn Bright Atomic Safety /401 0.L.
and Licensing Board Dockets 50-400 USNRC Washington DC 20555 George F. Trowbridge Shaw, Pittman, Potts & Trowbridge 1800 MN St. NW Washington DC 20036
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