ML19327A672
| ML19327A672 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 07/20/1989 |
| From: | Licciardo R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19327A642 | List:
|
| References | |
| TAC-73427, NUDOCS 8909140189 | |
| Download: ML19327A672 (77) | |
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Enclosure 3
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't AN EVALUATION OF THE CRITERIA FOR AND
'THE CALCULATION OF 0FFSITE DOSES DERIVING FROM-OPEN CONTAINMENT PURGE VALVES DURING A LOCA AT ZION UNITS 1 & 2 i
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DATED JULY 20, 1989 PREPARED BY ROBERT B. A. LICCIARDO REGISTERED PROFESSIONAL ENGINEER CALIFORNIA NUCLEAR ENGINEERING LICENSE NO. NU 001056 MECHANICAL ENGINEERING LICENSE NO. M015380 i
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INTRODUCTION On May 11, 1989, the writer submitted a memo on the subject:
DIFFERING PROFESSIONAL VIEW CONCERNING a)
Issuance Of SER to Zion 1/2 Allowing Full Power Operation With Open 42" Containment Isolation Valves.
b)
Methodology Used for Calculating Related Offsite Doses.
By memo of May II, 1989, from F. J. Miraglia to R. Licciardo, the writer was asked to clarify certain aspects of the regulatory positions used in his analysis including: a) Time to failure used in LOCA analysis and b) mechanisms for the transport of fission products from the primary (system) to the contain-
. ment. The writer was also asked to provide his view as to the safety significance of the Amendment roposed by management, and the safety significance of his concerns regardin LOCA analysis.
.This acterial was prepared in response to t ah t request and is in adjunct to his D.P.V which is attached to this document as Attachment 1.
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TABLE OF CONTENTS Pa2!
1 FISSION PRODUCT RELEASE FROM TUEL AND CONTAINMENT USED IN 1-1 ACCIDENT ANALYSES................................................
1.1 Radiological Source Terms Within the Core...................
1-1 L
- 1. 2 LOCA: Reg. Guide 1.4 Criteria:
Application to Zion.......
1-1 1.3 LOCA: BTP CSB 6-4 B5. Criteria............................
1-6 h
i 1.3.1 Characteristics of Fuel Failure Giving Fission Product Release During Postul'ated Accidents.........
1-8 1.3.2 Characteristics of Fission Product Released From Failed Fuel During Postulated Accidents........
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1.3.2.1 General....................................
1-12 1-13 1.3.2.2 Regulatory Guide 1.25......................
1-15 3.3.2.3 Regulatory Guide 1.77......................
1.3.2.4 Summary....................................
1-17, 1
1 l-1.4 BTP CBS 6-4 B5 Criteria:
Application to Zion...............
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2 0FFSITE DOSE CONSEQUENCES:
SUMMARY
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l 2-1 2.1 Basis for Calculations.....................................
l 2-2 2.2 Offsite Doses..............................................
l 2.2.1 RG 1.4 Source Terms Released Immediately on LOCA....
2-2 2-2 2.2.2 10% Gap Activity Released on DNBR...................
2-2 2.2.3 Equilibrium Gap Activity Released on DNBR...........
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-TABLE OF CONTENTS (Ccatinued)
Pagg 2.2.4 RCS @ 60 pC/gm. Activity:
Released to Containment
.lmmediately on a LOCA...............................
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2.2.5 RCS 9 60 pC/gm Activity:
Released Progressively to Containment on RCS' Discharge from a LOCA............
2-2 2.3 Conclusions.............:..................................
2-3 3
APPENDIX K EVALUATIONS, FUEL FAILURE AND FISSION PRODUCT RELEASE.
3.1 Preliminary...............................................
3-1 3.2 Review....................................................
3-2 3-5 3.3 Summary....................................................
4 CONCLUSIONS.....................................................
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l I FISSION "0 DUCT RELEASED FROM FUEL AND CONTAINMENT USED IN ACCIDENT ANALYSES 1.1 Radiological Source Terms Within The Core Exhibit I shows core and gap activities for Zion for iodine.
Calculated levels of iodine in the fuel clad gap are given to show a total I-131 EQU of 24.09 x 105 curies Total iodine in the core as I-131 EQU is 15.79 x 107 curies.
1.2 LOCA:
Rec. Guide 1.4 Criteria:
Aoplication to Zion Branch Technical ^ Position CSB 6-4 (Ref. 25) states that:
"The sizing of the purge lines in most plants have been based on the need to control the containment atmosphere during refueling operations. This need has resulted in very large lines penetrating the containetnt (about 42 inches in diameter).
Since these lines are normally the only ones pro-I vided that will permit some degree of control over the containment atmos-
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phere to facilitate personnel access, some plants have used them for con-tainment purging during normal plant operation.
Under such conditions, calculated accident doses could be significant.
Therefore, the use of these 1eroo containment purae and vent lines should be restricted to cold I
shutdown conditions and refuelina operations and they must be sealed closed in all other operational modes.
The desion and une of the purge and vent lines should be based on the premise of achievino acceptable calculated offsite radiological consequences and assuring emergency core cooling (ECCS) effectiveness is not degraded by a reduction in the containment backpressure.
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Purge system designs that are acceptable for use on a nonroutine basis dLrino normal plant operation can be achieved by providing additional purce lines.
The size of these lines should be limited such that in the event of a loss-of-coolant accident. assuming the purge valves are open and subsequently close. the radiological consequences calculated in accor-dance with Regulatory Guides 1.3 ~ and 1.4 would not exceed the 10 CFR Part 100 guideline values.
Also the maximum time for valve closure should not exceed five seconds to assure that the purge valves would'be closed before the onset of fuel failures following a LOCA.
Similar concerns apply to vent system designs."
This is interpreted by the writer as specifying that the large 42" purge and vent lines (PVLs) should be closed except in Modes 5 and 6.
And if purging is necessary in Modes 1, 2, 3 and 4, then smaller lines (8" and 10") should be considered and the source term to be used for evaluating 1'
offsite dose is that of Reg. Guide 1.4 which uses TID 14844 source terms as the fission product available for release to containment.
l RG 1.4.C Regulatory Position (Ref. 30) requires the following under related subsection No.:
"la. Twenty-five percent of the equilibrium radioactive iodine inventory developed from maximum full power operation of the core should be l
assumed to be immediately available for leakage from the primary L
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reactor containment. Ninety-one percent of this 25 percent is to be assumed to be in the form of elemental iodine, 5 percent of this 25 I
percent in the form of particulate iodine, and 4 percent of this 25 I
percent in the form of organic iodides."
i.e., 25% of the radioactive iodine inventory from exhibit 1 is specified to be immediately available inside primary containment for leakage to the l
atmosphere.
For Zion this would represent approximately 25 percent of curies of 1-131 EQU in the core i.e., 3.9 x 107 curies 7
15.79 x 10 immediately available inside containment for leakage to atmosphere.
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"Ic The effects of radiological decay during holdup in the containment or other buildings should be taken into account."
With half lives for iodine (1) varying from 3.16 x 108 secs for I-134 to 6.95 x 105 secs for 1-131, released immediately on a LOCA, and a time to valve closure of seven (7) seconds, there is no time for significant radioactive decay of any iodine isotope before it is discharged to atmosphere.
It is to be noted that the actual first stage of fission product release during a LOCA occurs with the infringement of DNBR for the fuel rod, leading to cherheating of the clad and fuel failure according to SRP 4.2 (Ref. 26) by perforation (or loss of hermeticity).
For Zion, this is specified to occur 0.1 sec's.into the event in the Appendix K evaluation l
of the LOCA event; the off-site calculations for this submittal have been made for a DNBR infringement of 1/2 sec. and are therefore less f
conservative.
"Id. The' reduction in the amount of radioactive material available for leakage to the environment by containment sprays, recirculating filter systems, or other engineered safety features may be taken into account, but the amount of reduction in concentration of radioactive materials should be evaluated on an individual case basis."
l During the first 7 seconds, there are no engineered safety features (ESF) j fission product clean up systems available for reducing fission product I
content prior to discharge to the environment.
Engineered safety feature containment sprays are initiated after 45 secs. Any filtration systems on l
the 42" inlet and outlet penetrations are not designed to ESF requirements.
Recirculating filter systems provided by W for fission product control of I
containment atmosphere during normal operations are not ESF equipment.
Containment volume of 2 million cubic feet originally containing 144,000 lbs of air reduces fission product discharged from the RCS by prior dilu-tion through mixing.
Exhibits 3 and 4, and 3A and 4A show the circumstances L
for containment and f.he discharging reactor coolant system.
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l The containment has an inttial mass of air of 144,000 lbs(atatmospheric j
pressure). On a LOCA, the initial rate of discharge from the RCS into containment is 75,000 lbi,/sec and over a period of seven (7) seconds prior to containment valve closure, a total of 270,000 lbs is so discharged.
This increases total mass in containment to 420,000 lbs, increasing total pressure in containment to'23.7 psig; at the same time a total mass of
'15,000 lbs [ valves fully open) to 2,860 lbs (valves partly open) of mixed i
containment inventory is discharged to the atmosphere.
i If it is assumed that all fission product released from the core is
.immediately available to containment as in RG_1.4, then total mixing of this product should be assumed to occur on initiation of the LOCA.
(The data presented show the results for a release second after the LOCA, but the differences are not significant for the intent of this submittal.) As a result, containment inventory discharged contains a L
uniform concentration of a decreasing curie content over the first-l 7 seconds, and the net result is a release to outside containment of 4.38%
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of the source term fission product inventory Q, released from the core on L
occurrence of the LOCA.
(A reduced amount of 1.57% is released for partly closed valves). Exhibit 2A shows that for the RG 1.4 source term, this l
gives a total release from containment over the first 7 seconds of 1.7 x 108 curies direct to atmosphere.
Related offsite dose is 490,000 rem for 2 x fully open valves.
Partially open valves reduce this to the value shown in Exhibit 2 of 612,000 curies and J56,000 rem.
e l-It should be recognized that the thermal-hydraulic, including energy conditions, are such that fluid is discharging from both the RCS and the i
l containment at very high energy levels, with associated pressure levels giving sonic discharge velocities into containment of the order of 1000 l
fps. Under these conditions it takes only hundredths of a seconds for RCS L
fluid to reach the containment isolation valves from the RCS system.
This is no comparison with the very low transport rates from the top of a fuel pool to conteinment isolation valves for a fuel handling accident inside 1-4 q
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J containment as discussed in Section 1.3.3.5 of this submittal; values of up sc.15' secs. have been considered appropriate for these circumstances.
If is assumed that the core fission product source term is instead uniformly l
mixed with the RCS Fluid prior to its dicharge to containment, (less con-servative than R.G. 1.4) curie content discharged to atmosphere is reduced from 4.38% Q to 1.9% Q where Q is the total term source released from the core by the LOCA and related source terms and related offsite doses are reduced by the same amount.
These are not unrealistic assumptions, for conservative purposes.
The LOCA causes sudden pressure drops in the RCS, to saturation pressures for I
the prevailing temperatures of the RCS inventory, causing steam release.
from violent boiling throughout the system.
This would cause substantial vibration of the fuel rods and movement of the prevailing damaged U02 pellets, facilitating the mass transfer'of fission product gases to and s
through the gap to the locally faulted cladding, followed by blowdown through the clad defects at high rates because of the prevailing pressure drops, between the gap and the core.
Over the first seven seconds of the event, heat is being tranferrred from the core to containment by steam formation at the core and subsequent mass transfer to the RCS system and break, and discharge to the containment, at the very high rates discussed earlier in this subsection.
Since fission product gases are released from the cladding, (and probably at the hottest sections) the transport of fission products released from the gap would be within the same steam and entrained liquid transport system to the break and then containment.
Within containment, unless special provisions have been made, there is no guarantee that a certain percentage of high concentrations of fission product inventory being released by RCS discharge is not being bypassed directly to the open containment isolation valves from its main path to principal containment volume.
In this sense, assuming an immediate release of all fission product to the containment on DNBR would help offset the potential non-conservatism of this bypass.
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"le. The primary reactor containment should be assumed to leak at the leak rate incorporated or to be incorporated as a technical specification requirement at peak accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> l
(0.1 percent per day), and at 50 percent of this leak rate for the l
remaining duration of the accident.
Peek accident pressure is the j
maximum pressure defined in the technical specifications for l
containment leak testing."
Except for dilution through mixing discussed under Id above, there is complete bypass of containment for 7 secs through the 2 x 42" open valves.
l 1he magnitude of discharge to the environment with related offsite dosrs l
has been discussed under ed above.
In reviewing these figures, it should be recognized that for a normal leakage of 0.1%/ day from containment, l
8 x 10 5% of Containment Inventory (Q), would be released in the same time frame of 7 seconds. When compared with 4.38%, this represents a dose l
reduction factor of 541.000 and wouid reduce the 7 second dose from 489,000 ren to 0.9 rem.
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Over a two hour time frame, and making allowance for 38 seconds without spray, followed by an iodine removal coefficient of $4/hr with a maximum i
reduction factor of 100, gives an approximate reduction in discharge by a factor of 32.000 leading to a ca).W.ted dose of 15 rem.
These reduction factors in offsite dose of 489,000 for the first seven l
seconds by effective early containment at 0.1%/ day, and of 32,000 in the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by effective containment at 0.1% per day and an iodine l
cleanup factor of 100, manifest the real significance of effective i
i containment and containment spray in fission product containment.
1.3 LOCA:
BTP CSB 6-4. 85 Criteria l
The Reg. 1.4 source terms of 1.2 above, are based upon the Regulatory requirement of 10 CFR 100.11, (a) footnote 1 (Ref. 36) that:
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"The fission product release assumed for these calculations should be based
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uponamajoraccident hypothesized for purposes of site analysis or i
postulated from considerations of possible accidental events, that would l
result in potential hazards not exceeded by those from any accident considered credible.
Such accidents have generally been assumed to result j
in substantial meltdovn of the core with subsequent release of appreciable quantities of fission products "
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However, Branch Technical Position CSB 6 4 (Ref. 25) provides another basis to justify containment purge design and which is less conservative than the l
Regulatory position.
This is given in related section B-5, as:
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The followino analyses should be performed to justify the containment l
purce system design:
r a.
An analysis of the radiological consecuences of a loss of-l coolant accident.
The analysis should be done for a spectrum of l
break sizes, and the instrumentation and setpoints that will l
actuate the purge valves closed should be identified. The source l
term used in the radiological calculations should be based on a calculation under the terms of Appendix K to determine the extent of fuel failure and the concomitant relasse of fission products.
I and the fission product activity in the primary coolant.
A pre-existing iodine spike should be, considered in determining primary coolant ac,tivity. The volume of containment in which fission productsaremixedshouldbejustified,andthefissionproducts
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from the above sources should be assumed to be released through j
the open purge valves during the maximum interval reouired for valve closure.
The radiological consecuences should be within 10 CFR Part 100 ouideline values."
To gain further regulatory interpretation of the meaning of fuel failure within this context, the writer's DPV (Ref 42) refers to SRP 4.2 FUEL SYSTEM DESIGN, I (AREAS OF REVIEW), 2nd para, (Ref. 26) which states that, in respect l
of postulated accidents:
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"The objectives of the fuel system safety review are to provide assurance f
that (4) the fuel system is not damaged as a result of normal operation f
and anticipated operational occurrences, (b) fuel system damage is never l
so severe as to prevent contral rod insertion when it is required.
(c) the number of fuel rod failures is not underestimated for postulated accidents, and (d) coolability is always maintained.
"Not damaged," as
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esed in the above statement, means that fuel rods do not fail, that fuel system dimensions remain within operational tolerances, anti that functional capabilities are not reduced below those assumed in the safety analysis.
This objective implements General Design Criterion 10 (Ref. 38), and the-design limits that accomplish this are called Specified Acceptable Fuel l
DesignLimits(SAFDLs).
" Fuel rod f ailure means that the fuel rod leaks and that the first fission product barrier (the claddina) has, therefore, j
been breached.
Fuel rod failures must be accounted for in the dose analysis required by 10 CFR Part 100 (Ref. 2) for postulated accidents."
The underscored lines show that fue rod failure in the context of this paragraph must be evaluated for postulated accidents and this evaluation must j
be conservative. Fuel Rod Failure means that the fuel rod leaks and that the first fission product barrier (the cladding) has therefore been breached; these failures must be accounted for in the dose analysis required by 10 CFR j
Part 100 (Ref. 36) for postulated accidents.
l Coolability is addressed as a separate criterion.
1.3.1 Characteristics of Fuel Failure Giving Fission Prodact Release During Postulated Accidents Regulatory clarification of fuel rod f ailure is given in SRP 4.2.II. A.2.
l (Ref 26) This is abstracted as follows for the circumstances of postulated accidents in particular:
"2.
FUEL ROD FAILURE
,This subsection applies to [nermai-eperatient-anticipated-operat(enai ocentrences--and) postulated-accidents.
[ Paragraphs-(t) through-(c)-address 1-8 i
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l f ailore mechanisms that ere mere iiniting dering normal operation;-and the inf ormatien-to be reviewed-shoeid-be-contained-in-Section-4;f-ef-the $efety
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Antiysis-Repert;) Paragraphs (d) throvoh (h) address failure mechanisms that
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are more limitino durino (anticipated operational occurrences and) postulated f
iccidents. (end the information-to be-reviewed-wiii essaily be contained in l
Enepter45-ef-the Safety-Antiysis-Report;"Peregraph-(43 shecid be eddressed I
in-$ection-4;f-ef-the Safety-Ametysis Report becesse-it-is not eddressed eisewhere)
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l To meet the requirements of ((aj-6enerai-Besign Criterion 16-es-it reintes-te Specified Acceptable Fuel Design Limits for normal operation, including antici l
pated operationai-occorrencest-and-(b)) 10 CFR Part 100 as it relates to fission i
product releases for postulated accidents, fuel rod failure criteria should be given for all known fuel rod failure mechanisms.
Fuel rod failure is defined i
as the loss of fuel rod hermeticity.
(Aitheogh we recognice-that-it is not j
pessibie-to-eveid-ali feei red f ailures and-that-eleenop-systems-ere insteiled j
to handie a-smali nember-of iesking-redst-it-is-the-objective-of the-review-to i
essure-that-feei does not-f aii-dee to-specific conses-dering-normei-operation and-anticipated-operationai-occurrtaces;) Fuel rod failures are permitted durino
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postulated accidents, but they must be accounted for in the dose analysis, i
Fuel rod failures can be caused by overheatino, pellet /claddino interaction l
(PCI), hydriding, cladding collapse, burstino, mechanical fracturino, and frettino.
Fuel failure criteria should address the followino to be complete.
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Only those failure mechanisms that are more limiting for postulated accidents are abstracted here:
(d) Overheatino of Claddino:
It has been traditional practice to assume that failures will not occur if the thermal margin criteria (DNBR for PWRs (and l
EPR for-BWRs)] are satisfied.
(The-review-ef these-criteria is-deteiied in SRP-Sec ti o n-4 : 4 ; ~ F or-no rmei-eperati on-a nd-anti ci pated-operati onai-oc c er-rencest-vietation-of-the-thermal-mergin-criteria-is-not permitted;] For postulated accidents, the total number of fuel rods that exceed the cri teria has been assumed to f ail for radiological dose cal:ulation purposes.
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i Although a thermal margin criterion is suf ficient to demonstrate the avoid-ante of overheatino from a deficient coolino_ mechanism, it is not a necessary condition (i.e., ONB is not a f ailure mechanism) and other mechanistic methods may be acceptable.
There is at present little experience with other approaches, but new positions recommendino different criteria should address claddino temperature, pressure, time duration, oxidation, and embrittlement.
(e) Overheatino of Feel Pellets:
(it hos else-been-treditionai practice te esseme-thet failore-eili ector-if-centeriine melting takes piecer" This ensiysis should be perfereed for-the meximon-iinear heet generation rate j
enywhere in-the core; incieding-sii het-spots-and-het-channei fseters; end shoeid-occeent for-the effects-of bornep end-composition-en-the menting l
pointr--Fer-normei operation-and enticipated-operationei ocentrences;
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center 44ne meiting is not permitted;) For postulated accidents, the total j
number of rods that experience centerline melting should be assumed to f ail for radiological dose calculation purposes.
(Thecenteriine-meitingcri-l terien wes estabiished to-assere that exisi-er-rediai reiocation of meiten i
feet wesid-neither siiew-melten-feel to come inte conteet with the ciedding f
nor prodoce-iecei het-spotst) The assumption that centerline melting results j
in fuel failure is conservative.
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(f) Excessive Fuel Enthalpy:
(Fer-e-severe reactivity initiated accident-(RI A) l in-e BWR st-tero-or-iew power; feei-fe49ere-is essemed-to-occer-if the-red 4-
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eliy-everaged-feti red enthaipy-is greate"r than if6-cei/g st eny exisi iece-f tion:) For full-power RI As in a BWR and all RIAs in a PWR, the tyrmal mar-l oin criteria (DNBR and CPR) are used as fuel failure criteria tu meet the ouldelines of Regulatory Guide 1.77 (Ref. 6) as it relates to fuel rod j
failure.
(The-ife-esi/g enthaipy-criterion in primarity-intended-te address-ciedding-overheating effects;-bet-it-sise-indirectly-eddress t
l peilet/ciadding-interactions-(Pti) ) Other criteria may be more appropriate l
for an RI A, but continued approval of (this-entheipy-criterien-and-the ther-mal marain criteria may be given until oeneric studies yield improvements.
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(g) Pellet / Cladding Interaction: There is no current criterion for fuel failure resulting from PCI, and the desion basis can only be stated generally.
Two l
related criteria should be applied, but they are not sufficient to preclude 1-10 l
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PCI failures.
(1) The uniform strain of the cladding should not exceed 1%,
(in-this conteatt emiform strain-(eisstie and-inetestic) is Wefined es transient indeced deformation with gege-lengths-corresponding to ciedding i
dieensionst steady state treepdown-and irrediation growth are-encindedt)
Although observing this strain limit may preclude some PCI failures, it I
wjQ1notprecludethecorrosionassistedfailuresthatoccuratlowstrains, nor will it preclude highly localized overstrain failures.
(2) Fuel meltino f
should be avoided.
The large volume increase associated with meltino may cause a pellet with a mniten center to exert a stress on the cladding, l
Such a PC) is avoided by avoidino fuel meltino, hotethatthissamecd l
terion was invoked in paragraph (e) to ensure that overheatino of the, j
cladding would not occur, j
i (h) Burstino:
To meet the requiremerts of Appendix K of 10 CFR Part 50 (Ref, f
- 9) as it relates to incidence of rupture durino a LOCA, (s reptere-tem-peratere-terreistion-mest be used-in the-tetA Ett$-enetysist) Zircaloy l
cladding will burst (rupture) under certain combinations of temperature, heating rate. and differential pressure. [Aitheagh-feti-suppliers may-ese f
different-reptere temperatere vs differentiai pressure-cervest-en-accept-
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able cerve shooid be-simiter-to the-ene-describe &in Ref 19 )
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t (i) Mechanical Fracturino: A mechanical fracture refers to a defect in a fuel t
rod caused by an externelly applied force such as a hydraulic load or a j
load derived from core plate motion.
Cladding inteority may be assumed if i
l the applied stress is less than 90% of the irradiated yield stress at the appropriate temperature. Other proposed limits must. be justifieo.
Results
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from seismic and LOCA analysis (Appendix A to this SRP section) may show that failures by this mechanism will not occur for less severe events."
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Summary:
Failure Mechanisms include:
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l (a) Infringment of DNBR criteria during postulated accidents which causes I
overheating of the cladding of the fuel rod, and is assumed to cause failure 1-11 i
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of the clad, and' release of contained fission products from the gap as a source term for the calculation of radiological doses.
(b) If postulated accident conditions cause calculated values of fuel pellet temperature to reach the melting point for the ura'iium dioxide at the centerline of the pellet, it is assumed that all such rods shall fail (and release fission products from the pellets - as well as the gap) for the calculation of radiological doses.
1.3.2 Characteristics of Fission Product Released from failed Fuel Durir.g Postulated Accidents 2.3.2.1 General Fission product release as source terms for postulated accidents relevant to the above fuel failure criteria are specified as:
SRP 4.2,Section I, last paragraph (Ref. 26) states that:
"All fuel damage criteria are described in SRP Section 4.2.
For those cri-teria that involve DNBR or CPR limits, specific thermal hydraulic criteria are given in SRP Section 4.4.
The available radioactive fission product inventory in fuel rods (i.e., the gap inventory expressed as a release fraction) is provided to the Accident Evaluation Branch for use in estimat-ing the radiological consequences of plant releases."
SRP 4.2.C.3(h) (Ref. 26) states that:
" Fission Product Inventory:
To meet the guidelines of Regulatory Guides 1.3,1.4,1.25 and 1,77 (Refs -6rf8-90) as they relate to fission product release, the available radioactive fission product inventory in fuel rods (i.e., the gap inventory) is presently specified by the assumptions in those -
Regulatory Guides.
These assumptions should be used until improved calcu-lational methods are approved by CPB (see-Refr-31)."
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Ine criteria from these Reg Guides are considered separately in the following subsections of this submittal in order to examine for general guidelines which
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may be applied to BIP CSB 6 4 B5 Criteria.
1.3.2.2 Regulatory Guide (RG) 3.25:
Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Hendling Accident in the Fuel l
Handling and Storage Facility for Boiling and Presserized Water Reactors j
i RG 1.25 (Ref. 31) covers the Fuel Handling Accident inside containment.
RG 1.25 p, age 25.1 under Section B, second para, provides for an immediate
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release of all activity from the fuel rod gap of the damage rods:
"The number and exposure histories of fuel assemblies assumed to be damaged determine the total amount of radioactive material available for immediate release into the water during a fuel handling accident."
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The same Section B fourth para, provides that:
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"Only that fraction of the fission products which migrates from the fuel l
matr!x to the gap and plenum regions during nomal operation would be avail-able for immediate release into the water in the event of clad damage.
I (Migration of fission products is a function of several variables including t
operating temperature, burnup, and isotopic half life taken into considera-tion ir, establishing the release fractions listed in this guide.)"
RG 1.25 also assumes that 10% of the total radioactive iodine in the rod (with calculated peak activity) is contained in the gap for release.
(See page 25.2,
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Item C.I.d):
j "All of the gap activity in the damaged rods is released and consists of 10% of the total noble gases other than Kr-85, 30% of the Kr-85, and 10%
i of the total radioactive iodine in the rods at the time of the accident.
{
1-13
l l
{
Released iodine rises to the surface of the related pool with a decoTtamination j
factor of 100, provided a minimum depth of 25 ft exists, and gap pressure is no l
greater than 1200 psig.
Subsequent treatment of the source term is typified i
i by the guidelines of SRP 15.7.4 Radiological Consequences of Fuel Handling Accidents (Ref. 28) which requires (under Section !!!.4, second and third para's that:
f I
"The reviewer should assess the time required to isolate the containment, f
This should include the instrument line sampling time (where appropriate),
detector response time and contsinment purge isolation valve actuation and closure time. The containment is considered isolated only when the purge isolation valves are fully closed.
The applicant's analysis should be reviewed regarding the travel time of any activity release starting fron its release point above the refueling cavity or transfer canal and l
including travel time in ducts or ventilation systems up to the inner containment purge isolation valve."
}
l "The time required for the release to reach the inner isolation valve is compared to the time required to isolate the containment.
If'the time required for the release to reach the isolation valve is longer than the time required to isolate containment, then essentially no release to the
)
l atmosphere occurs, and the reviewer's assessment should reflect this.
If the time required for the release to reach the isolation valve is less
'{
than that required to ist. late containment, and no mixing or dilution credit can be given, the reviewer should assume that the entire activity release escapes from the containment in evaluating the consequences.
Claims for credit for dilution or mixing of a release due to natural or j
forced convection inside containment are reviewed and assessed.
References (4) and [5] should be consulted and used by the reviewer for guidance in estimating dilution and mixing. Where mixing and dilution can be demon-strated within containnent, the radiological consequences will be reduced by the degree of mixing and dilution occurring prior to containment isolation."
1-14
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r l
l Related references (4) and ','J) are, i
i i
4.
Evaluation of Fission Product Release and Transport for a Fuel Handling Accident by G. Burley, Radiological Safety Branch, Division of Reactor Licensing, revised October 5, 1971.
l l
t 5.
Industrial Ventilation /A Manual of Recommended Practice - American l
Conference of Governmental Industrial Hygienists."
These circumstances relate to a set of containment environmental conditions in which mixing s.nergy is virtually absent, being provided by low energy contain-I ment purge and exhaust ventilation fans, and virtually no additional energy from the very small mass of fission product gas released from the damaged fuel elements, after travelling through a minimum depth of 23 ft.
Under certain conditions, this could provide for the total activity released (after decon-tamination in the pool) to be discharged directly to atmosphere outside
[
containment.
For Zion, the fundamental set of values for the th6tmal hydraulic parameters covering the above circumstances, are completely different to those governing j
the release and disbursement of fission products to the environment from a LOCA.
I 1.3.2.3 Regulatory Guide 1.77: Assumptions Used for Evaluating a control Rod EjectionAccidentForPressurizedWaterReactors t
Fundamentally, this Guide provides for an evaluation of the Thermal Hydraulic and Power conditions within the core, during the accident, to determine a) the extent of DNBR infringement and b) the amount of fuel exceeding the initiation
[
temperature of fuel melt (approximately 5150'F).
For Source Terms, RG 1.77, Appendix B1 (Ref. 32) proposes that:
"a.
The case resulting in the largest source term should be selected for i
evaluation.
l 1-1!,
l
)
f l
b.
The nuclide inventory in the fuel elements potentially breached l
should_be calculated, and it should be assumed that all caseous constituents in the fuel-clad caps are released.
l l
c.
The amount of activity accumulated in the fuel clad cap should be assumed to be 10% of the iodines and_10% of the noble cases i
accumulated at the end of core life, assumina continuous maximum f
full power operation, t
t d.
No allowance should be given for activity decay prior to accident 3
initiation, regardless of the reactor status for the selected esse, j
e.
The nuclide inventory of the fraction of the fuel which reaches or exceeds the initiation temperature for fuel meltino (typically 2842*C) at any time during the course of the accident should be I
calculated. and 100% of the noble cases and 25% of the iodine f
contained in this fraction should be assumed to be available fol f
release from the containment."
l Summarily:
The source term from molten fuel is the same as for RG 1.4.
The l
source term release from the gap is the same as for the fuel handling l
accident.
The subsequent effects of the release path on the ultimate source terms from containment are evaluated for each of two release paths, as if the other did not exist. These release paths are:
f 1
(1) By effectively immediate release of all source terms to containment to be
{
followed by the following cleanup and decay provisions which are the same as those normally accounted for in a LOCA in RG 1.4 (Ref. 30).
RG 1.77, l
App. B1 (Ref. 32) provides that:
i "f.
The effects of radiological decay during holdup in the containment or other buildings should be taken into account.
1-16 l
i l
l t
I
\\
g.
The reduction in the amount of radioactive m&terial available for f
leckage to the environment by containment sprays, recirculating l
filter systems, or other engineered safety features may be taken into j
account, but the amount of reduction in concentration of radioactive i
materials should be evaluated on a case by case basis.
i i
h.
The primary reactor containment should be assumed to leak at the leak
[
rate incorporated or to be incorporated as a technical specification j
requirement at peak accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at j
$0% of this leak rate for the remaining duration of the accident.
Peak accident pressure is the maximum pressure oefined in the f
technical specifications for containment leak testing."
i Additionally SRP 15.4.8, Section 111.3 (Ref. 27), furtt.er specifies that:
l "For releases via the containment building, 100% of the noble gases and 25% of the iodines contained in the fuel which is estimated to reach initiation of melting are assumed to be available for release from the f
containment."
Summarily:
For the release path to containment, these are effectively the
-[
provisions of RG 1.4 in respect of the treatment of Fission Product Source l
Terms after release from the core.
\\
(2) By release of fission products to the secondary system as per RG 1.77, Appendix B, Items li, j and k (Ref. 32).
i There are not considered in this submittal, as they do not apply to a release l
to containment.
i 1.3.2.4 Summary (of General Positions on Fission Product Releases Deriving frori RG 1.25 and 1.77)
(a) for failure of fuel cladding by either DNBR infringement or tuel handling accident:
1-17 j
- (
]
i l
[
For iodine, 10% of the fuel rod inventory is released from the gap.
For the control rod ejection accident this release is assumed to be available i
immediately inside containment for leakage.
I (b) For failure by centerline melting of the fuel pellet:
j l
25% of the iodine inventory of any fuel rod which reaches or exceeds the initiation temperature of fuel melting is assumed to be immediately l
available inside containment for release.
This is the same assumption applied in RG 1.4 for fuel melt deriving from a LOCA.
f 1.4 LOCA:
BTP CSB 6-4/85 Cri_teria:
Application to Zion l
I Zion fuel temperatures during normal operation at maximum power prior to a LOCA vary from 2500'F to 4100' for approximately 15% of the core (Exhibit 23). There will be a substantial increase in temperature of the whole core over a period l
I of up to 7 seconds following a LOCA and Exhibit 6 shows the related average i
cladding temperatures.
Considering the correlation of fission product release as a function of temperature shown in Exhibit 22, there is a high probability of a substantial increase in fission product activity in the gap over that of the equilibrium activity level represented on Exhibit 1, during these first j
seven (7) seconds of the accident, so that an increase in gap activity level from the equilibriua values shown in Exhibit I to the value of 10% used in the other postulated accidents is not an unreasonably conservative regulatory position to adopt for this event. On this basis, the iodine source term i
deriving from fuel rod failure by overheating of the fuel cladding by DNB infringement at Zion at 0.1 second into the event would be 157.9 x 105 curies l
cf I-131 EQU and is the value adopted by the writer in conformance to the related BTP.
In respect of fuel rod failure by centerline melting, the Zion i
FSAR (Ref. 33) does not provide detailed information on fuel pellet tempera-tures except for the general statement that the safety injection system prevents i
1-core meltdown Ref. 33, page 14.3-46 Revision 1 second para. ; provision for related fission product release from melted fuel rods is therefore not necessary for this evaluation to the guidance of the related BTP.
1-18 l
n i
I On the basis of BIP CSB 6 4, 85 therefore, a total iodine fission product l
release of 157.9 x '.05 curie 1131 EQU from the core, would be available to inside containment at 0.1 second into the LOCA.
By reference to the conditions r
inside containment discussed in detail in Section 1.2, items Id and le above, it can be shown that, the release of 157.9 x 105 curies of I-131 EQU from the core as a source term will result in the discharge of 692,000 curies of I-131 EQU to atmosphere with an offsite dose of 176,000 rem with 2 x 42" fully open
)
for 7 seconds, see Exhibit 2A, item 5.
With valves partly closed this is
{
reduced to 249,000 curies 1-131 EQU ar.d 63,400 rem, see Exhibit 2 item 5.
I It is noted that in its recent revision to the FSAR (Ref. 34 ) page 14.3 38 l
Revision 1.
h' has calculated an offsite dose from the LOCA on a non Reg. Guide 1.4 basis, by also using the entire inventory of fission products contained in the pellet cladding gap, sut has assumed the equilibrium values only, as listed in Exhibit 1.
This is equal to 24.09 x 1051-131 EQU which is 1.52% of the j
core activity as compared with the 10% exemplified in other NRC criteria and used by the writer.
Effective doses that would be obtained using equilibrium gap activity only are also presented in Exhibits 2A and 2 under items 4 and show offsite doses to thyroid are reduced to 27,000 rem for 2 fu11open valves and 9,700 rem for 2 partially closed valves, l
l I
b I
I D
1 1
1-19 l
t I
I 2 0FFSITE DOSE CONSEQUENCES:
SUMMARY
2.1 Basis for Calculations i
Based on discussions in section 1, radiological releases and related offsite f
contequences are shown in Exhibit 2A item 6 for 2 x 42" fully open (90') valves and Exhibit 2 item 6 for 2 x 42" values at a limited opening of 50'.
All calculations are based on valves closing in 7 seconds from commencement of 3
a LOCA.
Doses are based upon valves being in the open position for a full 7 j
seconds as required by the SRP.
Valves will be required by technical specifica-tions to close within seven (7) seconds of commencement of the LOCA.
^
For the sake of nample only, source terms are restricted to iodine in terms of 1 131 EQU, and thyroid dose only has been calculated. Dose is calculated at the site boundary (exclusion distance) of 415 meters.
Each dose is calculated l
independently of each other and are to be added to the LOCA leakage dose (over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) of 123 rem as appropriate.
An additional dose due to RCS inventory discharged into the containment would also need to be added, for all non-RG 1.4 calculations. These are given in Exhibits 2A and 2 under items 2 at 132 rem for 2x fully open valves, and 48 rem for 2 partially opened valves.
I For the diffusion coefficient, a value of 5 x 10
- sec/cm3 applicable to leakage conditions over a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period has been used.
In fact we have a high
[
L energy puff release of 7 seconds giving a potential finite cloud in travel to the enclosure boundary instead of a low leakage release diffusing into a cloud; I
as a result, the offsite dose under actual conditions.is likely to be increased.
for the 0-2 hour leakage, the licensee has used a more conservative value than the NRC of 9.2 x 10 4 sec/cm3 and this would increase dose by a factor of 1.84.
l l
l l
l 2-1 l
i
l
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t
t Exhibit 2A, item 6, shows that for fully (90') open 42" valves, the offsite f
dose for a RG 1.4 source term is calculated at 4B9,000 rem. And Exhibit 2, item 6, shows that for partially (50') open 42" values, these doses are reduced to 156,000 rem.
2.2.2 10% Gap Activity Released on DNBR Exhibit 2A (item 5) shows offsite doses reduced to 176,000 ren for fully open l
valves, and Exhibit 2 (item 5) shows reduction to 63,000 ren for partially open valves.
l l
2.2.3 Equilibrium Gap Activity Released on DNBR Exhibit 2A (item 4) shows offsite dose is reduced to 27,000 rem for fully open valves and Exhibit 2 (item 4) shows reduction to 9,700 rem for valves partially
[
open.
2.2.4 RCS 9 60 pc/gm Activity; All Released To Containment Immediately On A l
LOCA.
Exhibit 2A (item 2) shows offsite dose contrib'ution is 132 rem for fully open i
valves and Exhibit 2 (item 2) shows a reduction to 48 ren for partially open valves.
l This activity release is equivalent to DNBR infringement of only.08% of the l
fuel in the core.
l 2.2.5 RCS 9 60 pc/gm Activity; Released Progressively To Containment On RCS Discharge From A LOCA Exhibit 2A (item 3) shows offsite dose contribution is 58 rem and Exhibit 2 (item 3) shows a reduction to 21 rem for partially open valves.
2-2
i I
2.2 Conclusions (1) According to Reg. Guide 1.4 criteria the offsite doses are completely unacceptable.
(2) LOCA calculations for Zion show no fuel melt; however, for DNBR infringe-ment only, an evaluation of offsite dose based on release of 20% gap activity from 100% fuel still shows completely unacceptable circumstances.
Although this is in conformance with SRP 6-4, BTP, CSB B5 criteria, it is not in conformance with 10 CFR 100,11 (a) footnote 1 requirements which states that:
- The fission product release assumed for these calculations should be baseduponamajoraccident,hypothesizedforpurposesofsite analysis or postulated from conside' rations of possible accidental events, that would result in potential hazards not exceeded by those from any accident considered credible. Such accidents have generally l
been assumed to result in substantial meltdown of the core with subsequent release of appreciable quantities of fission products."
l (3) Partially closing the valve to 50' from 90' is not successful in reducing the offsite dose to acceptable values.
(4) With valves partially open at 50'; fuel failures by DNBR infringement on l
I a LOCA would have to be limited to 0.2% of the core to limit total doses l
to 10 CFR 100 limits, i
l l
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2-3
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l
)
i 3 APPEN0ix K EVALUATIONS, FUEL FAILURE, AND FIS$10N PRODUCT RELEASE
)
i 10 CFR $0.46 (Ref. 37), acceptance criteria for emergency core cooling system for light water nuclear power reactors, requires that during a LOCA, cladding l
temperatures, cladding oxidation, and hydrogen generation, are limited and such that the core remains amenable to cooling in the short run from the initial l
break through reflood, and also for long term post accident cooling.
t 10 CFR $0.46 does not include a requirement to evaluate the earliest time at f
which fission products could be released by local failure of the fuel cladding i
as fuel rod conditions rapidly change, challenge and exceed the limiting j
features of design which ensures fuel clad (and rod integrity) ender Normal i
Operating' Conditions and Transient Occurrences. These limiting features ar6 described as specified acceptable fuel design limits (SAFDLs) and are required l
under 10 CFR Part 50, Appendix A, Criterion 10.
l A principal feature of the Appendix K evaluation is that it is designed to
[
identify that rupture which causes a maximum post rupture cladding temperature l
within the fuel assembly being evaluated; and it is this time to rupture which is reported in the FSAR.
The Appendix K evaluation is not designed to report j
the earliest rupture that can occur.
3.1 Preliminary l
In evaluating 10 CFR 50.46 criteria through the use of the Appendix K evalua-tion model (Ref. 09), licensees are required to undertake a detailed evaluation l
of the items to be discussed below throughout the complete LOCA event, i.e.,
l from time 0 through 50-60 seconds, to determine that the clad rupture meeting l
the Appendix K criteria does not occur in the first 10-15 seconds of the event, l
and which is the region of special interest for this review.
In the time avail-l able for this research, a search of the UFSAR and the related reference mate-i rial on the docket does not disclose many of essential the details of this calculation (Ref's 1-24). We therefore use the limited information available to draw conclusions.
3-1 l
1
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l i
3.2 Review Appendix K calculations are undertaken on that fuel element assembly which ultimately provides the maximum clad temperature after (post) clad rupture.
Generic work by y (Ref. 17) proposes that maximum calculated temperatures
)
(post rupture) occurs in the low burn up (third region) fuel assembly.
]
Exhibit 6 shows the average clad temperatures deriving from Appendix K calcula-tions from the Zion FSAR, Figure 14 F. 2 19a, (Ref. 33). This shows that on infringement of DNBR at 1/10 second, average clad temperature increases very j
rapidly from a normal operating value of 720'F to at least 1350*F, and then to 1750'F, over a total period of seven seconds; thereaf ter temperature reduces g
rapidly to 1000'F at about 15 secs. from which it sharply increases ultimately j
to approx 2200'F.
l Exhibit 10showsthatyfuelsaredesignedtorequireayieldstrengthof l
45,000 psi a minimum for normal operations, and an ultimate tensile i
i strength of 57,000 psi as a damage limit, as specified acceptable fuel design l
limits (SAFDL).
Exhibit 11 shows that as temperatures increase above 850'F, j
l the available mechanical properties can be reduced below both these limits so l
that fuel clad cannot therefore be considered reliable in terms of protection L
against fission product release.
l l
Exhibit 10alsoshowsthatyfuelsrequireadesignlimitof1%oncladding strain as a design limit, and 1.7% as a damage limit. The work of this Sec-tion 3 will show how both these limits can be exceeded inside the seven seconds on infringement of DNBR during the course of a LOCA, so that again, fuel clad
[
cannot be considered reliable in terms of protection against fission product release.
l 1
Exhibit 15, shows how a temperature range of 1350'-1750*F traverses a range of l
Zircalloy metallurgical phases (transitions), a to (a + p) to p phases, during which ys = VTS and structural stability under stress is dependent upon mechan-ical/ strength properties which are a function of temperature and related time and stress at temperature.
Under the circumstance of the transient expected 3-2 I
l i
from Appendix K calculations with rapid changes of both temperature and stress, their is a need for empirical tests to determine swelling and burst (rupture) j characteristics under these same dynamic conditions.
Exhibit 15 represents l
results from such a series of tests (Ref. 13).
Such conditions are also represented in Exhibit 16 for Engineering Hoop Stress and temperature at rupture, for particular heating rates, and in conjunction f
with the information in Exhibit 20 on related rates of circumferential strain
,i on rupture, at the given rupture temperatures.
What are the expected operating pressure differentials across the clad under these LOCA conditions:
Reference information shows that internal clad pressure under normally operat-ing conditions is of the order of 1400 psig for new fuel and expected to l
l increase to 2250 psig at the end of the 3rd cycle (for the fuel). On this basis, we evaluate a gap pressure of 1500 psig at approximately 1/3 burnup into i
the first cycle, at which burnup maximum calculated clad temperatures are
[
expected on a LOCA.
l It is proposed that, immediately on a LOCA as clad temperature increases to l
1350'F, gap pressure will increase by 20%, to 1800 psig.
Exhibit 12 shows that 1
at this time, core pressure has reduced to 1500 psig giving a pressure drop across the clad of 300 psi which according to Exhibit 13 will give a hoop stress of approximately 2460 psi.
l At 7 seconds into the event, clad temperature has increased further to 1750'F, a total increase of 1030'F from the normal operating condition.
From this, it l
can be proposed that gap pressure for the complete rod can increase by 36% over its normal operating value to 2100 psig.
Exhibit 12 shows that at this time, i
core pressure has reduced to 950 psig so that the pressure drop across the i
clad is now 2100-950 i.e.,1150 psi which according to Exhibit 13 will give a l
l hoop stress of 9400 psi.
l When the above values cf pressure and temperature are plotted on a particular Hoop Stress vs Burst Temp curve (Exhibit 14) from reference 1, at one see the 3-3 l
~
l t
j i
i clad does not rupture, but at seven seconds the clad is well into the rupture l
- regime, f
i In its calculation of clad strain during Appendix K calculations, y uses results l
from tests by Hardy (Ref. 13).
Exhibit 15 is a set of results from one such test at 100'C/sec heat up rate (the heat up rate between 720'F and 1750'F in 7 seconds = 150F'/second (or 84C'/second)).
This exhibit shows that these Appendix K values over the first 7 seconds bracket the range from zero (0) expansion at 1350'F to the burst regime at 1750'F.
In respect to these values,
[
y has assumed that if clad strain reaches 10%, the clad will rupture; see Exhibit 18 from Ref. 3., Note that the SAFDLs of 1% and 1.7% on cladding strain can both be exceeded in the first seven seconds of DNBR infringement in the i
course of the LOCA.
The NRC, in its clad strain and rupture models uses the data shown in Exhibit
[
16 to determine when rupture is likely to occur for given rates of increase in temperature.
It is proposed by the NRC that the 28'C/S (=50F'/second) test j
points apply also to larger values (of rate of temperature increase).
l Exhibit 16 shows that the Appendix K values again bracket the complete set of l
experimental data and significantly at the higher temperatures of the transient.
)
Exhibit 20 shows the circumferential strain that can occur at given rupture temperatures, and the curve proposed by the NRC for Appendix K calculations.
Prime Facie; maximum strain gives maximum blockage leading to maximum calcu-lated temperatures for cladding after the burs't.
Infact,yhasestablished j
that maximum post rupture cladding temperature does not necessarily occur with a maximum circumferential strain at rupture, due apparently to direct radiation influences from fuel rods exposed by rupture at lesser values.
Providing rup-ture is expected by the data of Exhibit 16, the related strain is to be given by the NRC curve on Exhibit 20 (or lesser value giving maximum temperature).
It should be noted that with this information there would be a very high prob-ability of rupture at 1750*F down to 1500'F, with the probability decreasing, but still present at lower temperature.
Note that Exhibits 16 and 20 do show that fuel temperaturas and pressures could rupture the cladding over a whole range of conditions.
However, the purpose of 3-4 1
-.. ~ - -,.. _
,_.--...m_..
w
i.
i l
the Appendix K evaluation is to identify that particular rupture which would I
have the most conservative effect with respect to meeting the requirements of 10 CFR 50.46 and for this end, it models, and uses factors, to conservatively l
calculate values for the related parameters.
Its purpose is not to determine and identify when failure by bursting (rupture) first occurs as an otherwise evaluation of when fission product is first released. An example can be seen from Exhibit 16.
The test points can show marked deviations from what are apparently best estimate curves for the various rates of temperature increase.
t for conservatism in estimating the first occurrence of fuel rupture, one would have presumed the use of a boundary curve at the lower temperatures and pressures of each heating rate and Exhibit 20 would not have been required.
Note that Exhibit 15 does show that even though rupture may not occur with a l'
detailed re evaluation, cladding strain is most likely to exceed the 1% strain f
used by y (Ref 33 P. 3.2-39) as a SAFDL to meet the regulatory requirements of Ref. 38.
I The writer would be concerned about the relevance of the hoop stress, strain /
rupture data of Exhibits 16 and 20 to the power generation and heat trans-l fer conditions inside a reactor. These tests were done on electrically resist-ance heated cladding tubes. They do not simulate the heat transfer from central fuel rod pellets at high temperatures through a realistic gas gap of varying geometry, fuel pellet-clad contact, and pellet fracture / fragmentation l
to a cladding which is 12 ft long and which is,likely to have a much smaller l
l ratio of rupture length to clad length and gap volume than the test specimens.
The most revealing feature of Exhibit 16 is the data from the only test under-taken under much more realistic conditions, on a nuclear fuel rod using Zircalloy cladding in the TREAT reactor at ORNL; this information shows ruptures at very much reduced stress levels than the rest of the data.
i 3.3 Summary l
1.
Conditions within the core as currently evaluated by the Appendix K model, show that over the first seven (7) seconds following a LOCA, the following significant events occur:
3-5 l
o
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... _ _. - ~.. _.... -... _. _..,.. _.. - - -. - - - - - -. -..
L*
l i
1.1 DNBR for the whole core is infringed at 1/10 sec requiring gap j
activity at 10% core inventory for the whole core to be assumed as a source inside containment.
i
^
1.2 The temperature of the fuel clad, and the pressure drops across the j
same fuel clad, infringe specified acceptable fuel design limits l
(SADL) for normal operation and operational occurrences, required by 10 CFR 50 Appendix A Criterion 10.
Fuel rod failure must therefore l
be assumed for conservative calculations of offsite dose.
1.3 The temperature of the fuel clad and the related pressure drops show i
l conditions in which substantial deformation of the fuel clad by l
strain, can exceed the design and damage SAFDL values for cladding L
strain.
Fuel rod failure must therefore be assumed for conservative k
calculations of offsite dose.
1.4 The temperature of the fuel clad and the related pressure drops show j
conditions which could result in fuel rupture. This conclusion would l
needtobesubjecttodetailedverificationusingtheAppendixK I
model.
1.5 For Zion, fuel rods do not reach the melting point of the fuel pellets i
so that under minimum engineered safeguard conditions, additional fission product release from the fuel rods would not occur.
[
2.
The writer proposes that the purpose of Appendix K is to identify that particular rupture which would have the most conservative effect with respect to meeting the requirement of 10 CFR 50.46 and for this end it l
models, and uses factors, to calculate values for the related purposes.
l The purpose is not to determine and identify when failure by bursting l
(rupture) first occurs as an otherwise evaluation of when fission product j-is first released from the fuel summary a LOCA.
l t
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l.
l l
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6 3-6
~ _.
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1 f
4 CONCLUSIONS 1.
Conditions within the core as currently evaluated by the Appendix K model, show that over the first seven (7) seconds following a LOCA, the following significant events occur.
i 1
1 1.1 DNBR for the whole core is infringed at 1/10 sec requiring gap l
activity at 10% core inventory for the whole core to be assumed as a source inside containment.
)
i 1.2 The temperature of the fuel clad, and the pressure drops across the same fuel clad, infringe specified acceptable fuel design limits l
(SADL) for normal operation and operational occurrences, required by l
10 CFR 50 Appendix A, Criterion 10.
Fuel rod failure must therefore l
be assumed for conservative calculations of offsite dose.
1.3 The temperature of the fuel clad and the related pressure drops show conditions in which substahtial deformation of the fuel clad by strain, can exceed the design and damage SAFDL values for cladding strain.
Fuel rod failure must therefore be assumed for conservative calculations of offsite dose.
t 1.4 The temperature of the fuel clad and the related pressure drops show conditions which could result in fue.1 rupture.
This conclusion would' need to be subject to detailed verification using the Appendix K model.
- 1. 5 For Zion, fuel rods do not reach the melting point of the fuel pellets so that under minimum engineered safeguard conditions, additional fission product release from the fuel rods would not occur.
2.
The writer proposes that the purpose of Appendix K is to identify that particular rupture which would have the most conservative effect with respect to meeting the requirement of 10 CFR 50.46 and for this end it models, and uses factors, to calculate values for the related purposes.
4-1 l
i.
The purpose is not to determine and identify when failure by bursting (rupture) first occurs as an otherwise evaluation of when fission product is first released from the fuel summary a LOCA.
3.
As a result of the above 3.1 Fission product release from the fuel gap is a realistic considera-tion over the first seven seconds and prudent conservatism at this time should consider release from the whole core.
3.2 Reg Guide 1.4 deriving from Regulatory Requirement 10 CFR 100 requires consideration of substantial molten fuel as a design for the source term.
4.
The writer proposes that Regulatory philosophy recognized the possibility of Beyond Design Basis Events as the realism of a substantial commercial industry and therefore required protection against this occurrence and made provision in the Regulations for this purpose.
Considering the energy exchanges occurring in the core, and the insight of the Appendix K evaluations, it is not difficult to foresee significant fuel melt with potential additional substantive release of fission products from the fuel pellets over this time frame.
The question of the separate consideration of the timing of this additional contribution to the source term inside containment however must be moot. Uncontrollable release through open 42 inch CIVs is out of the question so that steps taken to correct that problem by effective isolation do resolve the unanswered philosophical question as to when fission products released by fuel melt should be more realistically and conservatively established.
4.1 A review of available fuel failure criteria, and the thermal-hydraulics aspects of the movement of fission gases from the clad to the environment over the first seven seconds of the event shows that:
4-2 l
f l
(a) The assumption of an immediate release to the containment is f
the only available conservative basis for use at this time, i
and that
[
[
(b) The physics of the large energy releases from the core clad through the RCS to containment, and through the open isolation
{
valves, shows effective mass transfer of fission product release j
from the clad to the environment within the same (7) secs.
i 5.
Fully open purge valves for a period of seven (7) secs. dischargs 1.7 x 108 curies of 1881 EQU to the environment giving an offsite dose of 489,000 rem to thyroid.
{
An isolated containment leaking at the safety analyses and TS limit of l
0.1% over 24 hrs, releases 3.14 curies of Ins EQU over the same seven i
seconds with a contribution to offsite dose of 0.9 rem.
The effectiveness of containment isolation and effective leak tightness in f
achieving a clean up factor of 541,000 over the first seven seconds of j
the LOCA is manifest.
{
6.
The offsite dose to thyroid for fully (90') open 42" valves using RG 1.4 l
source terms !s calculated at 489,000 rom. For partially (50') open 42" valves, these doses are reduced to 156,000 rem.
Reduction of source terms'
]
from RG 1.4 to 10% gap activity released on DNBR infringement reduces t
offsite dose to 176,000 rem for fully open valves with a reduction to 63,000 rem for partially open valves.
l l
Since the allowable limit for thyroid under 10 CFR 100 is 300 rem for 2 j
hrs at the Exclusion Boundary, these circumstances are unacceptable.
Therefore the 42" valves at Zion I and 2 should remain closed in Operational Modes 1, 2, 3 and 4.
7.
The stress / temperature relationships used to calculate fuel clad rupture to 10 CFR 50.46 are derived from test environments which are substantively non-realistic when compared with actual fuel rod conditions in a reactor 4-3 l
<=,
i l
daring a LOCA.
The only in reactor tests known to the writer at this time with the closest simulation of a real fuel condition gives ruptures at very much reduced pressures for given rupture temperatures. This comparison needs to be revisited to more thoroughly evaluate the reasons for the dif ferences and thereby improve our dwtailed knowledge of the total heat transfer environment which can lead to improvements in the calculational models of the fuel assemblies used in the Appendix K evalua-tions.
This can help in a improved definition of the limiting features of the circumstances and lead to ways and means of improving fuel clad design and performance for these circumstances.
e 1
l l
l 1
4-4
l j
g.
REFERENCES l
l 1.
Letter from E. P. Rabe (W) to James R. Miller (NRC):
Subject:
WCAP 9220-P-A.
j Rev. I " Westinghouse ECCS Evaluation Model 1981 Version."
l l
2.
Safety Evaluation Report on the 1981 Version of the Westinghouse Large Break ECCS Evaluation Model.
i 3.
Westinghouse Electric Corporation:
Locta-IV Program, Loss of Coolant l
f Tran'sient Analysis. Bordelon et al., WCAP8301. June 1974.
1 4.
Letter from E. P. Rahe (W) to M. Lauben:.
Subject:
Millstone Large Break Results for Increasing Fuel Burnup. July 20, 1981.
i S.
Letter from T. M. Anderson to J. Stolz:
Subject:
Additions to WCAP9220.
6.
Letter from T. M. Anderson (W) to J. R. Miller, dated May 15, 1981.
Proposed 1981 version of the Westinghouse Appendix K Evaluation Model.
7.
Rahe, E. P., " Westinghouse ECCS Evaluation Model 1981 Version," WCAP-9220 p-A Rev.1 (Proprietary), WCAP-9221-P-A Rev <1 (Non-Proprietary).
8.
USNRC:
Safety Evaluation Report on the 1981 Version of the Westinghouse Large Break ECCS Evaluation Model.
(
i L
9.
Letter from J. Stoltz (NRC) to T. Anderson (W);
Subject:
Safety Evaluation
[
of WCAP-9220, " Westinghouse ECCS Evaluation Model, February 1978 Version."
February 1978.
- 10. " Topical Report - Performance of Zircaloy Clad Fuel Rods During a Simulated Loss-of-Coolant Accident - Single Rod Tests," WCAP-7379-L, Volume I (Proprietary) and Volume II (Non proprietary), September 1969.
f 1
.l--
l
- 11. " Topical Report - Performance of Zircaloy Clad Fuel Rods During a l
Simulated Loss-of-Coolant Accident - Multi-Rod Tests," WCAP-7495-L, Proprietary, Volume. I - Test setup and Results; Volume II - Analyses of Results.
12.
Lorenz, Hobson & Parker. ORNL.
Fuel Rod Failure Under Loss-of-Coolant Conditions In TREAT.
Nuclear Technology, Vol II, May 1971-August 1971.
13.
D. G. Hardy: High Temperature Expansion and Rupture Behavior of Zircaloy Tubing, Atomic Energy of Canada Ltd.
Topical Meeting on Water Reactor Safety, Salt Lake City, Utah, March 26-28, 1973.
- 14. Letter from G. F. Owsley (EXXON) to J. R. Miller (NRC);
Subject:
Acceptance for Referencing of Topical Report XN-76-47(P), dated Nov. 5, 1981.
15.
D. A. Powers and R. O. Meyer, " Cladding Swelling and Rupture Models for LOCA Analysis," USNRC Report NUREG-0630, April 1980.
16.
D. H. Risher, et al., " Safety Analysis for Revised Fuel Rod Internal i
Pressure Design Bases," Westinghouse Report WCAP-8963-P-A, August' 31, 1978.
L
- 17. Letter from C. Eichelding (W) to J.F. Stolz (NRC) concarning " Review of WCAP-8963(P)," dated May 18, 1977.
- 18. Letter from T. M. Anderson (W) to J. F. Stolz (NRC); Subject WCAP-8963-P-A.
l.
Safety Analyses for the Revised Fuel Rod Internal Pressure Design Basis (Prop).
l l
l
- 19. Letter from J. F. Stolz, USNRC, to T. M. Anderson, Westinghouse,
Subject:
Safety Evaluation of WCAP-8720, dated February 9,1979.
20.
W. D. Leech, et al., " Revised PAD Code Thermal Safety Model," Westinghouse Report WCAP8720, Addenda 2, Octobe' 1982.
2
.]
1 l
\\_;
1 L
21.
Safety Evaluation of the Westinghouse Electric Corporation Topical Report WCAP-9500.
" Reference Core Report 17 x 17 Optimized fuel Assembly."
i.
May 1981; Core Performance Branch, Reactor Syster.s Branch.
22.
Letter from T. N. Anderson (W) to J. R. Miller (NRC);
Subject:
Proprietary Responses to the NRC request for additional information on WCAP-9500 (Proprietary).
January 22, 1981.
23.
Letter from Robert L. Tedesco, to Westinghouse Electric Corporation, Attention:
T. M. Anderson;
Subject:
Acceptanca for Referencing of Licensing Topical Report WCAP-9500.
- 24. WCAP-8785:
" Improved Analytical Models Used in Westinghouse Fuel Rod Otsign Computations."-
- 25. NUREG-0800 Standard Review Plan for the Review of Safety Analyses Report for Nuclear Power Plants.
July 1981, Section 6.2.4, Containment Isolation System, Branch Technical Position CSB 6-4, Contdinnent Purging During Normal Plant Operations.
- 26. NUREG-0800, Standard Review Plan, July, 1981, Scction 4.2, Fuel System Design.
- 27. NUREG-0800, Standard Review Plan, July 1981, Section 15.4.8.
Radiological' Consequences of a Control Rod Ejection Accident (PWR) Appendix A.
28.-
NUREG-0800, Standard Review Plan, July 1981, Section 15.7.4.
Radiological r
Consequences of Fuel Handling Accidents.
- 29. USNRC, NUREG-75/077, The Role of Fission Gas Release in Reactor Licensing, November 1975.
- 30. USNRC, Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors.
3 l'
l 1
- 31. USNRC, Regulatory Guide 1.25, " Assumptions Used for Evaluating the
}
Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors."
- 32. USNRC, Regulatory Guide 1.77, " Assumptions Used for Evaluating A Control 1
Rod Ejection Accident for Pressurized Water Reactors, May 1974.
33.
ZION 1 & 2: Updated Final Safety Analyses Report. Commonwealth Edison.
i
- 34. Deleted 35.
U.S. Covernment Printing office, " Reactor Design," Criterien 10, Appendix A. " General Design criteria for Nuclear Power Plants," Part 50, Title 10 Energy, code of Federal Regulations, January 1,1979.
36.
Ibid., " Reactor Site Criteria," Part 100.
- 37.. Ibid., " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors," Part 50.46.
38.
Ibid., " Reactor Design", Criterion 10, Appendix A, part 50.
39.
Ibid., Appendix K Evaluation, Part 50.
- 40. Memo from R. Licciardo (NRC) to T. E. Murley (NRC): Subject; Differing Professional View concerning (A) Zion 1/2 Containment Isolation Valves and (B) Nethodology Used for Calculating Related Offsite Doses.
Dated May 18, i
1989.
4 l
c'.
I f
EXHIBITS OF BACKGROUND INFORMATION RELATED TO DIFFERING PROFESSI0'NAL VIEW CONCERNING a)
Issuance of SER to Zion I/2 allowing full power operation with open 42" containment isolation valves.
t b)
Methocology used for calculating misted offsite doses.
4 l
f l
1 1
1.
+
0
.v l
l l
I 1.
t.
1 Tev1 ZION COREANDGAPACTIVITIES(IODINEONLY)
Operation at 3391 MWt for 500 days
. Assumptions:
Equilibrium Curies Percent-in the of Core Curies t
1 131 EQU inthg) Gap (X10 )
I 131 EQU Activity 5
Core 7) 7 (X 10 (X'10 x 10 in the Gap Isotope 1-131 8.35 8.35 2.3-19.2 19.2 1-132 12.75-46 0.26 3.3
.12 1-133 19.09 5.16 0.79 15.1 4.08 1-134 23.01
.39 0.16 3.8
.06 1-135 17.05 1.43 0.43 7.5
.63 7009 TF"79 P
l t
6 Y
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,___.+,7%_.,
.,.,.,e
,_v w-ww-w r
'T*w-
"w-
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==P r'fe-*"
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N
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)
1 Y
Rev. 1 J
i ZION: LOCA DURING CONTAINMENT PURGE USING 2x42" PENETRATIONS - VALVES OPEN 50' THYROID D0SE AT SITE BOUNDARY RESULTING ONLY FROM DISCHARGE TO CONTAINMENT DUTSIDE DURING CLOSURE i
(LOCA LEAKAGE DOSE (OVER 2 HRS) = +123 REMS)
{
Site /Excl.
Item Curies Discharged Boundary Dose i
No.
Source Radiological Sources I 131 E0 (Thyroid (REM)
I Licensee -1 131 EQ. 60 ue/gm in 73.5 18.7 RCS 50% cleanup in cont.
All released to r
containment on LOCA 2
RL I 131 EQ 60 uc/gm in 188 48 RCS, Allreleasedto
~
cont. on LOCA + 0.5 secs.
5
[ Total =0.119x10 curies) 3 RL I 131 EQ; 60 ue/gm in RCS. 82 21
~
l Released progressively to cont, with RCS discharge 4
RL 1 131 EQ; equiv gap 38,000 9676 activity (FgARcalc.)
[24.09 x 10 curies of I 131 EQ into cont.
onLOCA+0.5 secs.]
5 RL I 131 E0; SRP Gap activity 248,950 63.400 at 101 Total Activity 5
(SRP calc.) [157.9 x 10 curies of I 131 EQ into cont. on LOCA + 0.5 secs.)
1 l
6 RL I 131 EQ; Reg. Guide 1.4 611,500 155.700 at25%TogalActivity l.
[390 x 10 curies of 1 131 EQ into cont. on L
LOCA) 1 l'
X = 5 x 10'# sec/m3 for 0-2 hrs. at annimum exclusion distance of 415 meters
[NRC]U
[ Licensee has used 9 x 10'" sec/m3 for SARs)
~
x 3-i 1
?
t v
l ZION 1 ft 2 h
CONTAINMENT INVENTORIES DURING LOCA BLOW DOWN c
U
- nCS me.. Di. heres note
/
Into,Contelnment r
9 Cumulative Discheree of J
RCS Into Containment 400 x 10s 100 s
[
& Cumulative Mass of Air
.E l
.i and RCS Discherge
[
ah i
)
U r
l 300 x 108 M f
I y
E i
f
{
L ('
i I
i s
T g
3m i
p
\\j[
200 x we j s0 i 8
u a
{
40 g
JL
{
ai
'k 30
\\
100 x 103 30 t
h w A b
- w 10 3 *% m ll l T
0 "O
0 4
8 12 16 30 24 3B I'
Time After Break - Seconds
\\
L
x 4
j ZION 1 ft 2
'i CONTAINMENT' THERMAL ~ HYDRAULIC CONDITIONS FISSION PRODUCT INVENTORIES a x 42" unos valves open only a0*
Instead'of 90' Fully Open -
At 7 Secs 154,406 Lbs Air 272,100 lbs RCS 428,820 Lbs Press a 23.70 psig I
f
.R ' '
Fission Product Inventory
= 0.h4 x Q Released at 0.5 seos Discharge Rates
. Cumulative Totals Discharged Air + RCS inventory j
Air + RCS inventory 1023.08 Lbs/sec B379 Lbs
(.237% inv.)
4-f j+
Fission Product inventory Fission Product inventory
.237% Q/sec 1.588% of Q i
(Q = Fission Product Inventory Released at t = 0.5 seos) f
\\
e
s E
i FISSION PRODUCT DISCHARGED TO OUTSIDE CONTAINMENT l
II EFFECT OF ASSUMPTIONS ON FISSION PRODUCT RELEASE TO CONTAINMENT 2 x 42" lines.
l Valves open 50' t
Given Q = total inventory of fission products in RCS at T=0.5 secs after LOCA If Q is released instantaneously to the total containment volume:
Fission proouct inventory discharged outside containment over 7 secs = 1.5685 Q If Q is released over ties with RCS inventory and based on a uniform distribution within the inventory:
(
Fission product inventory discharged outside containment over 7 secs = 0.5615 Q i
c W
e e
r
( I.
f
d h.
I 2A Rev. I l
ZION: 'LOCA OURING CONTAINMENT PURGE USING 2x4?" PENETRATIONS - VALVES FULLY OPEN -(90')
THYROID DOSE AT SITE BOUNDARY RESULTING ONLY FROM J
DISCHARGE TO CONTAINMENT OUTSIDE DURING CLOSURE (LOCA LEAKAGE DOSE (OVER 2 HRS) = +123 REMS)
H Site /Excl.
1 Item Curies Discharged Boundary) Dose (REM)
(Thyroid No.
Source Radiological Sources I 131 EO 1
Licensee I 131 EQ. 60 uc/gm in RCS 204.3 52
~~
50% cleanup in cont.
All released to containment on LOCA 2
RL I 131 EQ, 60 ue/gm in 522 132' RCS. All released to cont.
=059 b5 curies]
o L
3 RL I 131 EQ; 60 uc/gm in RCS.
227 58 i
Released progressively to cont. with RCS discharge l
4 RL I131EQ;equivgapactigity 105,600 26.878 (FSAR calc.)
curiesofI1[24.09x10 31 EQ into cont.
on LOCA + 0.5 secs.]
5 RL I 131 EQ; SRP Gap activity 691,520 176.010 at los Total Activgty (FSAR L
calc.)[157.9x10 curies L
of I 131 EQ into cont on L
LOCA+0.5 secs.]
6 RL I 131 E0; Reg. Guide 1.4 1,698,592 488.911 at25%TogalActivity
[390x10 curies of I 131 EQ into cont. on o
LOCA]
X = 5 x 10'# sec/m8 for 0-2 hrs. at minimum exclusion distance of 415 meters
[NRC)U
[Licenseehasused9x10'4 sec/m3forSARs]
I
-.~..-r..
. ~. - ~.. -.
..--. ~..+
a 3A
- 1....
I J
i J
ZION 1 ft 2 CONTAINMENT INVENTORIES J
DURING LOCA BLOW DOWN b
- RCS Mass Discharge Rate
/
Into Containment
$ Cumulative Discharge of r
RCS Into Containment 400 x 103 J
L d
& Cumulative Mass of Air
.E O 100 8
and RCS Dischstge g3$
90 u
7 I
a so y,
g 300 x 103 i
i f
U' 2
,/
70 '
g
[
N.-
)1 i
=.x 10 e
.0 3
=
l
\\l t '
1 40 s
J L
(
g i
I b gg 100 x 108 h,g l
a e g
\\%
- =mw l
,n in w
% m L
qp
%0 0
4 8
12 if 20 24 23
[
Time After Break - Seconds
\\
L
- '-- ~ m
v u
ZION 1 ft 2 n.
s S
. CONTAINMENT THERMAL HYDRAULIC CONDITION I
FISSION PRODUCT INVENTORIES.
3 1..
<o 2 x 42" Lines Fully Open
. At 7 Seu
'i 154,480 Lbs Air 262.474 Lbs RCS i
~
416,834 Lbs
. Press a 23.79 psig f
Fission Product inventory g' '
= 0.966 x 0 Reloosed e
at 0.5 secs Discharge Rate Air + RCS Inventory
(
Curnulative Totals Discharged 2000 Lbs/sec Air + RCS Inventory
( 882% inv.)
15026 Lbs t
~
Fission Product inventory Fission Product inventory
.882% Q/sec 4.38% c,f Q l
(0 = Fission Product inventory Released at t = 0.5 secs) l
(~
t l
.y
.y, e
's FISSION PRODUCT DISCHARGED TO OUTSIDE CONTAINMENT EFFECT OF ASSUMPTIONS ON F1SSION PRODUCT RELEASE TO CONTAINMENT P x 42" lines fully open (90').
Given Q' = Total inventory of fission products in RCS at T=0.5 see after L If Q is released instantaneously to the total containment volume Fission product inventory discharged outside containment over 7 secs = 4.381 Q If Q is released over time with RCS inventory, and based on a uniform f
distribution within the inventory:
Fission product inventory discharged outside containment over 7 secs = 1.901 Q
.c W
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TIME (SECHDS) l b
A M A D / K M C A. 6 R $ f 7 V N U N C k
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(Unit 1)
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3.1.3.3 Thermal and Kydraulic Limits The reactor. core is designed to meet the following limiting thermal and hydraulic criteria:
l The minimum allowable DNBR during] normal operation, including l
a.
anticipatedtransients,is[1.30*.
b.
No fuel melting during any anticipated operating condition.
To maintain fuel rod integrity and prevent fission product release, it is necessary to prevent clad overheating under all operating conditions.
This' is accomplished by preventing a departure from nucleate boiling (DNB).
DN8 causes a large decrease in the heat transfer coefficient between the fuel rods and the reactor coolant resulting in high clad temperatures.
f
[
t i
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{
l l
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\\
8 The integrity of fuel rod claddin so as to retain fission products or fuel i
material is directly related to e adding stress and strain under normal
\\,'
operating and overpower conditions. Design limits and damage Ifmits (cladding f
perforation) in terms of stress and strain are as follows:
Damage Limit Design Limit Stress Ultimate strength Yield strength-57,000 psi minimum 45,000 psi minimum
. Strain 1.75 1.05 The damage limits given above are minimum values. Actual damage limits depend upon neutron exposure and normal variation of material properties and would generally be greater then these minimum damage limits.
For most of the fuel rod life the actual stresses and strains are considerable below the design limits. Thus, significant margins exist between actual operating conditions and the damage limits.
l The other parameters having an influence on cladding stress and strain and the relationship of these parameters to the damage limits.are as follows:
U 1.
Internal gas pressure:
The internal as pressure required to produce cladding stresses equal to the damage li it under normal operating conditions is well in excess of 8-the maximum design pressure. The maximum design internal pressure under
\\..
nominal conditions is 2250 psia which is equal to the coolant pressure.
The end of life internal gas pressure depends upon the initial pressure,
.A void volume, and fuel rod power history, however it does not exceed the design limit of 2250 psia.-
N 2.
Cladding temperature:
l The strength of the fuel cladding is temperature dependent. The minimum L
ultimate strength reduces to the design yield strength at an average l
cladding temperature of approximately 850'F. The maximum average i
l cladding temperature during normal operating conditions is given in Table 3.2.2-1[as720'F].
l' 1
l f
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l
i
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S, bl a
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'revious experience with removaole rods has been attained at Santon, Yankee l
and'2orita; and soditional experience will be acquired at the San Onofre Cycle 2 and Surry Unit 1.
Over 300 fuel rods were removed and re-losetted into assemelles during the Saxton re-constitution without evidenceiof failure.
Leak detection tests were performed on the assemblies after all rods were re-inserted, and no leakage was detected. An equally large nwC" of Saxton rods have been successfully removed, examined and re-inserted i&, over 12 3x3 subassemolles at Saxton. In addition, 28 full length Yankee rods were removed, examined and re-inserted into Yankee Core y special assemelles.
Similiar handling of 22 removable rods was sucessfully completed during the first Zorjta refueling. All such fuel handlings have been cone routinely sne without difficulty.
The same fuel roc de' sign limits indicated in section 3.2.3 fuel tt :tratu:e ano internal pressure, are maintained for these rewvable roes and *here is n:
reduction in eargin to DNS. Their inclusion in the initial 21:n Unit I c<.. e 1:adin int ccu:es no seditica.a1 safety co siderations ano in r: o t y :.13 s the ss!epunte snelyses and triateo engineering infcrmation cresente: in
- eviously submitted material in sup: ort of the license a:elicMI:n.
i 3.2.3.5 Evaluation of Core Components Fuel Evaluation
[
./
The fission gas release and the associated buildup of internal gas pressure in I
((
the fuel rods is calculated by a code based on experimentally determined o
rates. The increase of internal pressure in the fuel rod due to this phen:mena is incluoed in the determination of the maximum cladcing stresses at the end of core life when the fission product gap inventory is a maximum.
The maximum allowable strain in the cladding, considering the comsined effects L
of internal fission gas pressure, external coolant pressure, fuel pellet L
swelling and clad creep is limited to less than 1 per cent throughout core life. Tne associated stresses are below 'the yield strength of tne material uncer all nermal ocerating conditions.
To assure that manufactured fuel rods meet a high standard of excellence from
~
the standpoint of functional requirements, many inspections and tests are l
performed both on the raw material and the finished product. These tests and l:
ins:e:tions include chemical analysis, elevated temperature, tensile testing l-of fuel tubes, dimensional inspection, X-ray of both end plug welds, I-ultrasonic testing and helium leak tests. See additional details in Section
- 3. 3. 3.1.
I In the event of cladding defects, the high resistance of uranium di xide fuel cellets to attack by hot water prote:ts against fuel deterioration or ce:: East L
in fuel integrity. Thermal stress in the pellets, while causifg some fracture of the bulk material during temperature cycling, oces not res:dt in pulverization or gress void formation in the fuel matrix. As dshown by ocerating ex erience and extensive experimental work in the industry, the
{\\
thermal esign parameters conservatively account for any changes in the ther.tal cerf:rmance of the fuel element due to cellet fracture.
0115T 3.2 38 0060A
~[i 4
]
1 l
'Q,
' The consequences of a breach of cladding are greatly reduced b'y the ability of l
uranium cloride to retain fission products in:Juding those which are gaseous or highly volatile. This retentiveness cecreases with increasing tem:erature and fuel burnvo, but remains a significant factor even at fulf. power coerating temperature in the maximum burnup element.
A survey of'high burnup uranium dioxide" fuel element behavior indicates that for an initial uranium dioxide void volume, which is a function of tre fuel censity, it is possible to conservatively define the fuel swelling as a function of ournwo.
The fuel swelling model cons!:ers the effe:t of eurnup, tem:erature cistribution, and internal volds.
It is an entitleal n::el..ni:n 5ss :een :ne: a: =1th :sta from 5ettis, vaneee, CvTA, Saxton 6n: f.
1 s.
Tne
.oellet censities for the tnree regions are listed in Tatle 3.2.3-1.
Tne integrity of fuel rod cladding so as to retain. fission predxts or fuel material is cirectly related to c3 adding stress and strain unter r.::*al
- eratin; an: Overo:mer con:!tions. Cesi;n limits and ta.a;e lic.'.ts 1:la::ir; oerferation) in terms of stress and strain are as follows:
Damace Limit -
Desion Limit i
Stress Ultimate strength Yield strength-l 57,000 psi minimum 45,000 psi minimum i
Strain 1. 715
- 1. 05 The damage limits given above are minmlum values. Actual dama upon neutron exoosure and normal variation of materjal propertfe limits depend
~
it es and would L
generally be greater than tnese minimum damage liftito. For most of the fuel rod life the actual stresses and strains are t.omeiderab'y below the cesign L
L'
. limits. Thus, significant margins exist between actual operating conditions p
'and the damage limits.
t l.
The other parameters having an influence on cleading stress and strain and the relationship of these parameters to the damage limits are as fc12cws:
, 1.
Internal gas pressure:
^
The internal gas pressure required to produce cladding stresses equal to the damage limit under normal operating conditions is well in excess of the maximum design pressure. The maximum design internal pressure under l :'
nominal conditions is 2250 psia which is equal to the coolant pressure.
l The end of life internal gas pressure depends upon the initial pressure, i
vold volu' e, and fuel rod power history, hcwever it does not exceed the m
i l
design limit of 2250 psia.
l-2.
Cladding tem:erature:
u The strength of the fuel cleading is temperature ce;endent, The minimum i
ultimate strength reduces to the cesign yield strength at an average
/
cladding tem:erature of accroximately 950*F.
The maximum everage :1 acting
((
tem:erature during normal oce: sting concitions is given in Ta 1 3.2.2-1.
0115T 3.2 39 0060A r
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. 5
' data shown in Figure _bl.
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THE ROLA 0F Piml0N GAS RELEASE IN REACTOR LICENSING.
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1 NOVEMBER 1975 i
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I ZION 3
CORE TEMPERATURE DISTRIBUTION r
k Assumptier,st Operation at 3391 MWt for 500 days l
a h
1 of Core Fuel Volvow Local Temperature. 'F Above the Given Tessierature
)
4100 l
0.0 3700 O.2 3300 l
2.8 2900 7.0 2500 14.5 l
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Attachment UNITS 0 sTATss
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- i NUCLEAR REGULATORY COMMISSION
.{
meneeworon. e, c. seems g "
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May 11, 1989 Thoses 1. Murley MEMORANDUM FOR:
Office of Nuclear, Director Reactor Regulation i
FROM:
Robert 8.A.Licciarde,ReactorEngineer(Nuclear) l Plant Systems Branch Division of Engineering and Systems Technology l
SUBJECT:
DIFFERING PROFE5$10NAL VIEW CONCERNING a)
Issuance of $ER to Zion 1/2 allowing full power i
operation with open 42" containment isolation j
- valves, j
b) Methodology used for calculating related offsite doses.
l The writer submits a Differing professional View (DPV) in accordance with the provisions of NRC Manual Chapter 4125.
l This issue has arisen out of the Safety Evaluatfon Report (SER) undertaken for the Zion Units 1 and 2 as prepared by the writerl see Attachnert.
l
?
The principal issue is the prudent and conservative calculation of the additions I f to offsite dose which may result from a LOCA at a facility during the use of l
4 open purge supply and exhaust valves at full power.
The licensee for Zion 1/2 Was proposed full power operation of the facility I
with the 42' purge supply and exhaust containment isolation valves open to a 11mited position of 50', and capable of isolation within seven (7) seconds of the cossencement of a LOCA.
i The writers $tt concludes that the 42' valves at Zion should remain closed in Modes 1, 2 3 and 4 because the conseevenca of the offsite dose to thyroid i
(from fodine),during a LOCA is unacceptably highu whole body has not been i
evaluated. The least value for the additional offsite dose which may be proposed within the licensing basis is 64,000 rom over the first seven (7) j seconds of the LOCA. Management staff has diseereed with the writer's methodology and conclusion and plans issuance of a separate SER permitting i
l the operation requested. The writer reevests non issuance of the related SIR i
to the Itcensee. He also proposes probability of a generic action on other facilities which have been granted such licenses based on the staff's current pathodology.
L in general, t'he management staff has adopted a criterion described in $RP BTp CSB 6-4 which is that providing the maximum time for closure of these L
containment isolation valves does not exceed 5 seconds (and by plant specific exception, up to 15 seconds), then the valves would be closed before the onset of fuel feilure following a LOCA so that the only contribution to offsite dose l
'/
is from RCS operational levels of fission product directly dischgyggd into
('
containment during this period, and then through the open containment isolation valves before closure.
-gievyqui.
-~--__ _ _ ___
y t
-2 Thomas E. Murley
/,/ r l
g In evaluating the consequence for Zion, the writer has used an alternata Criterion in BTP CSS 6-4 which states that.
'The following analyses should be performed to justify the cont,ainment purge system design:
An analysis of the radiological consequences of a loss-of-coolant The analysis should be done for a spectrum of break Sites, and the instrumentation and setpoints that will actuate the accident.
i The source tem used in l
purge valves closed should be identified.
.the radiological calculations should be based on a calculation under i
the terms o Appendfx K to detemine the extent of fuel failure and l
o the concomitant release of fission products, and the fission productA p; i
activity in the primary coolant.
The volume be considered in determining primary coolant activity.
i of containment in which fission products are mixed should be i
justified, and the fission products from the above sources should be j
assumed to be released through the open purge valves durino the The radiologueal maximum interval required for valve closure.
i consequences should be within 10 CFR Part 100 puideline values.'
i l
Using these related guidelines for Zion,(the fuel performance over t 1s detailed and shows that fuel failure by infringement of DNBR criteria) f occurs within i seconds of the comencement of the LOCA, and together with ot licensing basis responses including fission product release from the fuel gap U'
containment and discharge
\\
and the thermal hydrevlic conditions in the core, ion products to the l
I nozzle, result in a substantive discharge of fiss l
environment of far greater consequence than are calculated by the staff.
i The relative consequences of these differing approaches are that whereas the staff methodology gives additions to offsite dose resulting in total doses within 10 CFR Part 100 timits, the alternate approach used by the writer shows a substantially increased offsite dose exceeding 10 CFR part 100 limit i
l with completely unacceptable consequences to Public Health and Safety.
l i
The writer reevests review of the Differing Professional View in a timely manner in accordance with the pr visions of NRC Manual Chapter 4125.
- fM Robert 8. A. Licciardo Registered Professional Engineer California Nuclear Engineering License No. NU 001056 Mechanical Engineering License No. M 015380 J. Sniedek cc:
- 0. Muller S. Varga
/
C. Patel se i.
F. Miraglia L. Shao i
A. Thadani J. Wermie)
J. Kudrick
I 7
e UNITS 0 sTATss
\\
NUCLE AR REGULATORY COMMIS$10N
(
a l
massessestow, s. c. seres May 11, 1989 i
1 Docket Nos. 50 235 Attachment and 50 304 MEMORANDUM FOR: Daniel Muller, Director i
Project Directorate !!! 2 Division of Reactor Projects !!!, !Y, Y l
F and Special Projects i
i FROM:
Jared 5. Werniel, Acting Chief Plant Systems Branch Division of Engineering and Systems Technology i
i
SUBJECT:
0FF5!TE RADIOLOGICAL CONSEQUENCES OF LOCA DURING j
CONTA! *ENT PURGE PROPOSED IN T5 CHANGES FOR ZION 1 AND 2 Reference LettertoN.R.Denton(NRC)FromP.C.Leonarddated February 2,1946,
Subject:
Zion Nuclear Power Station, Units 1 and 2 Proposed Amendment to Facility Operating License No. DPR 30 and DPR-48 1
Plant Name:
Zion Nuclear Power Station, Units 1 and 2 j
Licensee Cosmonwealth Edison Company l
(/
TAC Nos.:
55417 and 55418 Review Status:
Couplete i
k i
Zion Units 1 and 2 (Ceco) has responded to an NRC request to propose 75 to f
primarily constrain operation of the large (42') containcent purge supply and exhaust valves on these units see reference 1.
i i
L
.The former Plant Systems tranch, Section A, of the Division of PWR Licensing l
A, requested Section 8 of the same branch to review the offsite radiologica l
consequences of this proposal.
l The enclosed Safety Evaluation Report has been prepared by the technical reviewer t
initially assigned to this task, namely Robert B. A. Licciardo.
l sal is to allow fv11 power operation of the facility with The Itcensee's pro and exhaust containment isolation valves open to a the 42" purge supp limited position o 50' andcapableofisolationwithinseven17)secondsof j
the cossencement of a LdCA.
l t
The review concludes that the 42' valves at Zion should remain closed in Modes 1, 2, 3. and 4 because the consequence of the offsite dose to thyroid during a LOCA is unacceptable hight whole body dose has not been v
(from iodine)he least value for the additional offsite dose which may be proposed evaluated: T withinthelicensingbasisis64,000removerthefirstseven(7) seconds.
The convehtional treatment of BTP C5B 6 4 which assumes that fuel failure does [
not occur over the first 5-15 seconds after a LOCA and thereby t) t'entpr.ly RC5 and operating inventory of fission products is released to the contfinalt analyses for containment response, and licensing basis requirements (including criteria) for the calculation for, and the occurrence of, fuel damage and the quantification and treatment of resulting source terms.
'a 1
i i
)o l
Caniel Muller 2
I
)
/,/ r
\\
Our SALP input is provided in Enclosure 2.
We consider our efforts oe TAC l
Nos.-55417 and 56418 to be complete.
)
l Jared 5. Wemiel. Acting Chief Plant Systems Branch Division of Engineering and Systems Technology i
Enclosures:
l As stated l
i cc w/ enclosures j
C. Patel i
CONTACT: R. Licciardo 370876 t
h
+
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Daniel Muller 2-
.y r
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Our SALP input is provided in Enclosure 2.
We consider our efforts on TAC Nos. 55417 and 55418 to be complete.
i Jared 5. Werniel, Acting Chief Plant Systems tranch Division of Engineering and Systems Technology l
7 Er. closures As stated cc w/entissures:
i C. Patet CONTACT: R. Licciardo X20876 i
F57A"SUT!0N j
mene" ruas t
Plant File
[
JWersiel l
\\.
JKudrick i
l RArchittel AThadani L5heo
-l TGody ($ ALP only)
~
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RLicciardo i
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l SPL8: DEST SPLB: DEST
$PLB: DEST RLicciargotcf JKudrick JWerniel l
5////89 5/ /89 5/ /89 f
sa 5
5520 NAME: Zion TACs 55417/8 Licciardo 1
l
\\'
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UNITS 0 sTATSs I
NUCLE AR REGULATORY COMMITS 10N t
(
q i
seAemmetow. e. s. senes j
)
e<
8 SAFETY EVALUATION BY THE OFFICE OF WUCLEAR REACTOR RESULATLON I
PLANT SY$ FEM $ BRANCH i
0FF$!TE RADIOLOGICAL CONSEQUENCE OF LOCA DURING CONTAINMENT PURGE UNITS 1 AND 2 ZION NUCLEAR POWER STATION,d 50 304 DOCKET N05. 50 235 an
1.0 INTRODUCTION
ZionUnits1and!(Ceco)hasrespondedtoanNRCrequesttoproposeT5to primarily constrain operation of the large (42') containment purge supply and enhaust valves on these units.
The former Plant Systems Branch, Section A, of the Division of PWR Licensint A, requested Section 8 of the same branch to. review the offsite radiologica consequences of this proposal.
2.0 EVALUATION l
[
tac % ground review she'ws that the facility was evaluated on the basis of 5,
normally closed punt valves so that these consequences were never included in the Zion SER. Further, that a IrMar from Westinghouse JW) to Cosmonwealth LOCA and Containment Purge * (Ref.1976 en the subject of 'Of31te 00 Edison Compasty dated October 22 2)hasneverbeenevaluatedbytheNRC.
Subsequent to that TN!.2 event, the operability and automatic control of these valves was evaluated leading to the request for the required 75 Radiologic 3)whichwas 1
intended to be resolves in a subsequent probabilistic risk assessment which definitively excluded it from consideration without esty justification (Ref. 4).
uses an RC$
The W analyses undertaken under Cassenwealth Edison instruction,f the accident operitional inventory of 60 oc/gm equivalent i 131 at the time o with a resulting site boundary thyroid dose due to iodine (during closure of the valves), of $2 rom, and which added to the containment leakage dose of !!3 rem gives a total 176 rem which is within the 10 CFR 100 limit of 300 rom.
The total todine inventory of the RCS is assumed to be released into containment on initiation of the LOCA: a los plate out is assumed leaving the residual los as part of containment inventory for discharge out through both fully open containmentpurgelinesforatotalofseven(7 seconds).
However, when reviewed against the BTP C58 6-4, Item 8.5.4 requires thatt
'The source term used in the radiological calculations should be based on a caleviation under the teres of Appendix K to determine the entent of L
/
fuelfailureandtheconcospitmentreleaseoffissionprodjuts.andthe j
i fission product activity in the primary coolant."
tt l
N e
.t.
i
'? !
i Further: $RP 4.2 identifies fuel failure with infringement of DN8R criteria, with the related requirement that gap activity be considered as part of the source ters, and Regulatory Guide 1.77 recossnends that under similar circumstances, gap activity should be assumed at 105 of core activity.- Fuel damage criteria also includes the occurrence of center line relting'With measures of addttional activity release also guided by Regulatory Guide 1.77, but the Zion SAR shows this does not occur.
RevisingthesourceterstoAppendixKcalculations(inwhichallfuelgoes l
to DNBR in i second) with related release of all gap activity into containment, with limited blowdown to offs'te during the related 7 seconds closure time and absent a los plate out of iodine as can be interpreted from the above referenced item B.5.a. increases offsite dose due to containment purge above rem and would thereby be completely unacceptable.
by a factor of 3400 to 176 000 Limiting the purge line valves to en opening of 50' could reduce offsite dose tn 64,000 res and represents the least value which may be proposed within the Itcensing basis.
Note: The BTF C58 6-4 proposing that valve closure within 5 seconds will ensure purge valves are closed before the onset of fuel fa11eres has since been extended by the s+eff on a plant specific basis to 15 seconds. Further, the writer cannot find any safety evaluation report Supporting these positions.
l These positions cannot be sustained for Zion since a) DNBR infringement (from l
AppendixKcalculations)andhencefuelfailureandgapactivityrelease[Ref.
(t Ref. Regulatory Guide 1.77) occur 'sithin i
$RP 4.2) of 105 of core inventory (A, b) related maximum clad temperatures of l
second of the initiation of the LOC l
(
1750'F occur issnediately and never reduce below 1400'F c) RCS pressure in the l
region of the core rapidly reduces from 2250 psia to gb0 psia in 7 seconds I
increasingpotentialpressuredropacrossthecladdingforreleaseofgap I
d)themassivebulkbelingandblowdown I
activity to the RCS inventoryltimately discharges 270,000 lbs of RCS inventory surrounding the failed fuel u into the containment at 7 seconds into the event increasing containment pressure from 0.3 psig to 23.8 psig (in these 7 seconds), and e) causes 15,000 lbs of the resulting containment inventory to be discharged to the environment through 2x42" fully open lines, or 6400 lbs for the same lines with valve closed to 50'.
l l
3.0 CONCLVS10N The 42' valves at Zion should remain closed in Modes 1, 2, 3, and 4 because the consequences of the offsite dose to thyroid (from iodine) during a LOCA is unacceptably highs whole body dose has not been evclusted. The least value for offsite dose to the thyroid which may be proposed within the existing licensing basis is 64,000 res.
t The conventfeeal treatment of BTp C$8 6 4 which assumes that fuel failure does not occur over the first 515 seconds after a LOCA and thereby that only RCS 4
operating inventory of fission products is released to the containment, and then to the environment, cannot in general be sustained against thermal hy(draulicincluding analyses for containment response, and licensing basis requirements
. (.
criteria) for the calculation for, and the occurrence of, fuel d pagg and the A
quantification and treatment of the resulting source terms L
e 0
l l
i ae References
\\
l 1.
LetterfromP.C. Blond (CECO)toH.R.Denton(hRC); Subject Zion, Units 1 and 2 Proposed Amendment to Facility Operating'L4 cense Nos. DPR-39 and DPR.44 dated Februtry 21, 1986.
2.
Letter from R. L. Kelley LW) to C. Reed (CECO);
Subject:
Offsite J
Dose During LOCA and ContITnment Purge, dated October 22, 1986.
j 3.
LettertoL.O.DelGeor9e(CECO) frees.A.Varga(NRC):
Subject:
Generic Concerns of Purging and Venting Containments, dated September 9, Ital.
l
" Evaluation of the Risk at Zion," dated August 14,194I.
Subject:
l Memo for F. H. Robinson from R. W. Nouston 4
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$PLB $ ALP INPUT Plant Name:
Zion Nuclear Ger. orating Stations, Units 1 and 2
$tR subject: Containment Purge and tent Valve Operation TAC Nos.:
55417/8 l
l Suunary of Review /insoection Activities The licensee provided an evaluation of offsite doses undertaken in 1976. This was undertaken with a methodology and source ters chosen by the licensee. The l
licensee did not present results from alternative more detailed methodologies which could be considered enforceable under existing regulatory positions and the related circumstances.
l Narrative Discussion of Licensee Performance - Functional Area The single only methodology used by the licensee is not an acceptable approach for est< mating doses under the proposed circumstances and especially since alternate detailed evaluations required by the SRP give greatly increased values beyond 10 CFR Part 100 limits. A prudent approach would have recognized the deficiencies and risks in the single methodology adopted with resulting substantively different reconnendations to ensure public asalth and g/*
safety.
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Author: Robert 8. A. Licciardo i
Date:
May 11, 1989 i
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