ML20011D094

Comments on Plant Containment Isolation Valves,Per SAD-89-54 Request

ML20011D094
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	08/08/1989
From:	Kudrick J Office of Nuclear Reactor Regulation
To:	Thadani A NRC
Shared Package
ML20011D095	List: ML20011D094 ML20011E949 ML20011E952 ML20011E953 ML20011E956 ML20011E958 ML20011E959
References
NUDOCS 8909140138
Download: ML20011D094 (2)

Text

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i ENCLOSt1RE 2 i

August 8, 1989 i

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k PCTE TO: Ashok Thadani SLEGECT: MV CDORJ1tG CURAltiTNT IEG.ATION W1.'ES AT ZICH Fur your request in SAD 89-54, I have reviewed ths cfplac6b(11ty of Reg.

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Guldet. EC's and BTP's cited in tre NV.

Tre follcwit>g ark my rcrmonto cri the subject.

4 its major reference within thrr NV that is with6n tre EPJ' uccoo is Branch Tactitical Fosition CSD 6-4.

This DTP Is referented jo Eif metation 6.2.4 Cmtainmmt Isolation System. It>arver, the faces of trw WV only addresses the contents of BTP. To present a complete picture of tlw staf f's position, I tulieve it is worttwhile to note the elements c4 SFF h.T!.4 <rd W the BTP CSB 6-4 is referwnced.

FP 6.2.4, ACHFT4G CRITERIA does provide some guidarco in this regard.

j Specific criteria necessary to meet the relevant recNirements of the regulations for purge valves is provided 'in subesction n.

First of all, the guidance for closure time states For lires which provide an open path from the ccritaiment to the envirans; e.g., the containment surge and vent lines, isolation valve clonure tines "en tiw order of" 5 seconds or less may be necessary.

Note that the intent nust be taken as a goal but does not preclude closure times grunter than 5 seconds.

It also refers to BTP 6-4 for further guidence.

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i O ita fir.o1 referetice in the ETF is made which is relevant to the isste at hwd.

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Thet. is t.tn refermee to tre need to perforvii dose 6iinalysis. Subsection n l

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...rixiarding tie site of tte purge system used during normal plant opwaticri wid tic.lustification by acceptable dose consequence analysis, may be wiwd if tte applicalt commits to limit the use of the p.irge system to less than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year while the plant is in thu startup, power, tot stardby and hot shutdow1 modes of i

operaticris.

Thew added rwferences more properly reflect the staf f view on purging.

It does not irdicate that the staf f during the dowloptient of the !FP believed that tin consequences of purging at the time of a LDCA would result in the impact asserted in the IFV.

furytrid these carmenits, I believe that the DPV cited the correct secticris of the BTP.

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ENCLOSRE 3 l

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NOTE TO:

Ashok Thadant FROM:

Ted Quay

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SUBJECT:

DPV CONCERNING ZION PURGE OPERATION Without going over each specific item in detail, Mr. Licciardo's DPV stems f rom his misapplication of an instantaneous release of fission products to the containment (and to the environment) through the open purge valves.

The LOCA Regulatory Guides (1.3 and 1.4) were intended to establish the source term to be used for evaluation of the containment design f eatures f or mitigation of the radiological consecuences of the LOCA.

The " instantaneous" source term used in accidents such as LOCA was established to ensure containment isolation features incorporated either fast acting valves or features that would ensure that containment integrity was not compromised during operation (e.g.,

dual doors on personnel locks).

Use of the instantaneous source term prevented a case-by-4 case analysis of each containment incorporating slow isolation features.

This simplified approach was never intended to be applied to purge valves except for those purge systems that i

incorporated extremely slow closing valves.

No opening in containment could be justified using the instantaneous source term.

Consecuently, no purge / vent system design could be f ound acceptable using a simplified instantaneous release assumption. Without purge vent capability plant operations would be tremendously restricted.

The SRP recognized the lack of realism of the " instantaneous" release assumption and established that l')

purge / vent systems needed extremely f ast acting valves (and in most cases, this means air operated so that power failure results in closure without reliance on any ac/dc power source such as the diesels); and 2) a more realistic source term, maximizing the fission product content of the primary coolant during opacational conditions, (spike coolant activity for a PWR-60 uCi/gm) should be employed.

A. number of years ago, Limerick had a two inch " vent" line that had extremely slow acting isolation valves (closure time on the order of two minutes).

NRC imposed the instantaneous source term on the evaluation of the doses from this line for two reasons. The first is that for a 2" line there is no reason 0.01 to incorporate a fast acting valve.

Second, the risk from failure to isolate the containment increases the longer a line is left open due to LOCA produced debris.

The Zion DPV analysis argues that substantial core damage occurs fast.

The information provided by Wayne Hodges disputes this. Complete core damage does not occur instantaneously and even assuming the fast releases of substantial amounts of fission products, the DPV ignores progression of fuel damage with heatup and consequently, any transport time to the break.

The first pound of primary coolant released to the containment (and for

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2 that matter to the environment) is assumed to contain the same amount of fission Droducts as a pound released from the primary system much later in the blowdown phase.

Taking this p,gument to the extreme, all the fission products would be releared during the blowdown phase which in some cases could ts u es short as ten seconds.

Although the SRP specifies 5 seconds, AEB accepted closure times up to 15 seconds based on inf ormal discussions we had with Research on their severe accident analyses.

We were told that even up to 20 seconds that no substantial releases would occur.

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i NOTE TO: Frank Miraglia FROM:

Ashok Thadani l

SU8 JECT: DPV CONCERNING CONTAINMENT ISOLATION val.VES AT ZION I

The attached note from the Reactor Systems Branch provides revised L

analyses to recognize higher linear heat generation rates for Zion. The conclusions, however, provided to you in my note dated August 11, 1989, remain unchanged, b

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Enclosure:

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i AUG 2 31989 NOTE TO:

Ashok Thadani, Assistant Director for Systems Division of Engineering A Systems Technology Office of Nuclear Reactor Regulation FROM:

M. Wayne Hodges, Chief Reactor Systems Branch Division of Engineering & Systems Technology Office of Nuclear Reactor Regulation 1

$UBJECT:

DPV CONCERNING CONTAINMENT ISOLATION VALVES AT ZION On August l0, I sent you a note concerning the subject DPV with an attached memorandum from Norm Lauben.

Recently, revised, and higher linear heat generation rates were obtained from Zion for pins with burnups greater than 40,000 MWD /MTU. As a result of this new information, Nom Lauben has performed a revised analysis which is given in the attached memorandum.

My review of the new analysis shows that the conclusions given in the previous note remains valid. This is, rupture of high burnup fuel pins during the blowdown transient is not credible for existing fuel designs.

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M. 'Wa e Hodges, Chief Reactor Systems Branch l

Division of Engineering & Systems Technology Office of Nuclear Reactor M gulation

Enclosures:

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A Chief MEMORANDUM FOR:

M. Wayne Hodges, Branch Reactor Systems Division of Engineering & Systems Technology Office of Nuclear Reactor Regulation FROM:

G. N. Lauben, Section Leader 7

Accident Management Section Reactor & Plant Systems Branch Division of Systems Research Office of Nuclear Regulatory Research

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$USJECT:

COMMENTS ON A DPV CONCERNING EARLY BLOWDOWN CLAC' DING AUPTURE DURING A LARGE BREAK LOCA Per your request, I have reviewed certain aspects of the DPV on Containment Isolation Valves at Zion.

In particular, I addressed the issues raised with respect to cladding rupture of high burnup high pressure fuel early inblowdownpriortocontainmentisolation(about7 seconds). The comments are enclosed. If you have ser questions, please contact me on x23573.

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_. $ n G. N. Lauben, Section Leader Accident Management Section Reactor & Plant Systems Branch Division of Systems Research Office of Nuclear Regulatory Research

Enclosures:

As stated cc: R.B.A Licciardo A. Thadani

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l Coments on a DPV Concerning Early Blowdown Cladding Rupture During a large Break LOCA

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inaDPV(Reference 1)BobLicciardohaspostulatedthatPWRfuelrodswith high burnup and high internal pressure could sustain cladding rupture within a few seconds of a large break LOCA prior to containment isolation. This is further postulated to lead to large off-site releases. Following is some l

information which may be helpful in addressing some of the issues in the DPV.

Seven issues in the DPV are first addressed, then some preliminary observations are made. The DPY issues are referenced by page number and a quote or sumary of.the issue.

Issue 1 (p. 3-1)

" Appendix K evaluation is not designed to report the earliest rupture that can occur."

(Also,seepp.3-4and3-5.)

While Appendix K does not specifically require searchir.g for the earliest rupture, early ruptures would always be the worst with respect to 50.46 limits if they were calculated to occur. Vendor analyses in the past have shown that because of the extensive cladding swelling prior to rupture, the resultant low transient gap conductance severely limits blowdown heat removal. As a consequence, vendor evaluation model calculations showed that the 2200*F PCT was always exceeded. Therefore, the vendors would always need to reduce the peak power to avoid early blowdown cladding ruptures. Vendor steady state fuel thermal performance and subsequent LOCA analyses showed that the peak linear heat generation rate (PLHGR) was always low enough to avoid early blowdown swelling and rupture for high burnup pins. These studies were done about 13 to 15 years ago with Appendix K evaluation models which are no longer used. I do not know if analyses with high burnup pins have been done with recently approved fuel performance and LOCA models. The older analyses always I

showed that low burnup post densification pins were always most limiting, in fact, because the PLHGR was highest and gap conductance was very low. High The burnup pins are lowest in PLHGR although the pin pressure is highest.

combination of high cladding temperature and higher internal pressure are needed to cause cladding rupture.

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Issue 2 (p. 3-2)

"This shows that on infringement of DNBR at 1/10 second, average clad temperature increase very rapidly from a nomal operating value of 720*F to at least 1350*F, and then to 1750'F, over a total period of seven i

seconds."

1750'F is indeed a very high early blowcown peak cladding temperature (PCT),

but virtually impossible for a high burnup pin with a much lower PLHGR.

If a high burnup pin reached 1750'F, at 7 seconds it would most likely rupture.

More realistic LOCA analyses have been performed as part of the Code Scaling, l

Applicability, and Uncertainty program in RES. A best estimate analysis was performed and code uncertainties evaluated for a large break LOCA (Reference 2).

In order to accomplish this, sensitivity studies were perfomed which varied gap conductance, peaking factors, and several other variables. The plant used was a Westinghouse 4-loop 3411 MWt plant with 17x17 fuel and a low f

burnup of only 16000 MWD /MTU which resulted in a PLHGR of 9.35 kw/ft. The blowdown peak for the nominal CSAU case was 1103'F (see Figure 1). Based on over 250 clad temperature calculations and using Monte Carlo sampling techniques, it was determined that the 95th percentile blowdown PCT was 1447'F.

It has been determined that 15x15 pins (as used at Zion) with burnups greater than 40,000 MWD /MTU have PLHGRs no greater than 6.4 kw/f t.

Using the CSAU

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calculated sensitivity of blowdown PCT to LHGR, the value of 1447'F can be-extrapolated to approximately 1320*F for the 6".4 kw/ft PLHGR high burnup 15x15 pin. This illustrates that the 1750*F blowdown PCT calculated by Westinghouse is quite conservative, especially for a high burnup pin.

I believe that this Westinghouse calculation is probably at least 10 years old.

Issue 3 (p. 3-2)

' Exhibit 10 also shows that }[ fuels require a design Ifmit The of 11 on cladding strain as a design limit, and 1.75 as a damage limit.

work of this Section 3 will show how both of these limits can be exceeded inside the seven seconds on infringement of DNBR during the course of a LOCA,

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3 As exhibit 10 states, these design values are for nominal operation or overpower conditions, g LOCA. Also, DNBR infringement has never been j

considered the operant criterion for fuel failure during a LOCA. Although, I am told that this is not as clear as it should be in the standerd review plan or any applicable regulatory guides.

Incidentally, PBF LOCA test do not show DN8 occurring until 3-4 seconds for a very severe LBLOCA (Reference 3).

Issue 4 (p. 3-3)

"...there is a need for empirical tests to determine swellingandburst(rupture)characteristicsunderthesesamedynamic conditions."

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The results of the PBF LOCA tests satisfy this condition and will be discussed l

as part of !ssue 7.

J Issue 5 (p. 3-3)

" Reference information shows that internal clad pressure under normally operating conditions is of the order of 1400 psig for new fuel and expected to increase to 2250 psig at the end of the 3rd cycle (for the fuel)."

It is not known what reference information is being invoked here. GAPCON calculations show the following results.

I G9 TABLE 1 GAPCON Pin Pressure Calculations Code fuel PLHGR Burnup Pressure kw/ft MWD /MTU (psig)

GAPCON 15x15 15 0

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The PAD The Reference 4 GAPCON calculations were performed 9 to 10 years ago.

3.4model(Reference 5)wasapprovedbytheNRCfordesignandsafetyanalysis in May 1988. Proprietary calculations done with PAD 3.4 showed substantially lower pressures at comparable burnups and PLHGRs.

It is well known that the GAPCON fission gas release model is very conservative. The PAD calculations were done at an arbitrarily high PLHGR and would show an even lower pressure j

at'the reduced kw/ft.

Issue 6 (p. 3-3)

"It is proposed that,1emediately, on a LOCA as clad temperature increases to 1350'F, gap pressure will increase by 205, to 1800 psig.... At 7 seconds into the event, clad temperature has increased further to 1750'F,.... From this, it can be proposed that gap pressure for the complete rod can increase by 36% over its normal operating value to 2100 psig."

The basis for concluding that pin pressure increases during an LBLOCA blowdown is not known and contrary to the evidence. A series of 31arge break LOCA simulations (Reference 3) (LOC-3, LOC-5, and LOC-6) were performed in P8F with well instrumented Zircoloy clad 00, fuel elements pre-pressurized to simulate low and high burnup PWR fuel. PBF blowdowns are quite severe compared to postulated PWR LBLOCA blowdowns.

In PBF, the pressure decrease and rate of l

mass loss is very rapid. No good reverse flow' blowdown heat transfer is evident as is the case in LOFT results or PWR analysis. Figure 2 (Reference 6)showsthefuelrodpressureforrod3intestLOC-3. Also, shown are FRAP-T6 calculations using two different plastic deformation models. Clearly, pressure decreases throughout the transient.

Figure 3 is a plot showing measured pressure decrease for Rod 11 in Test LOC-6. A FRAP-76 characterization calculation was done for a postulated LBLOCA in Zion (reference 7)whichalsoshowedapressuredecreasethroughoutthetransient.

Issue 7(p.3-5)-Concernisexpressedabouttherelevanceofelectrically heated rods used in defining the swelling and rupture curves in NUREG-0630.

ItissuggestedthattheTREATdatashowninNUREG-0630(Reference 6)wouldbe more realistic. Also, on pp. 4-3 and 4-4, this concern is restated.

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pressure versus time for Rod 3 of PBF test LOC-3.

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It is clear that TREAT data is anomalous compared to the electrically heated rods and is attributed to difficulties in obtaining accurate temperature data I

in the burst region. A better source of in-reactor data is the PBF series l

discussed previously. Figure 4 is a plot from NUREG-0630 (Reference 8, j

Exhibit 16).

Included are data points with temperature uncertainty for the 9 ruptured rods in the PBF LOC series of tests, and the FRF data from TREAT.

It is clear that the more recent P8F data is very consistent with the NUREG-0630 curves.

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Observations Regarding L8LOCA Blowdown Rupture of High Burnup Fuel Rods.

The main contributor,s to fuel cladding rupture,are high pressure drop across the cladding and high cladding temperature. Early post-DNB cladding temperatures are determined to a very large degree by pre-accident stored enery which is a function of local peak power (PLHGR), pre-accident gap conductance, effective U0, thermal conductivity, blowdown heat transfer, and critical flow model. TheCSAUstudy(Reference 2)confirmedthisassessment.

Of these variables, only PLHGR is controllable by plant operators, and then only to a limited degree. High burnup, third cycle fuel is always placed in low power regions. Pin pressure is determined by pre-pressurization and fission gas release. As shown in References 3 and 6, pin pressure does not exhibit,a direct functional relationship to blewdown cladding temperature.

Asnotedearlier,theCSAU17x1795thpercentilePCTof1447'F(Reference 2) l The could be approximately extrapolated to 1320'F for a high burnup 15x15 pin.

15x15 PCT calculated at 13.26 kw/ft (Reference 7) was 1543'F. The Zion hot pin did not rupture in Reference 7.

The Reference 7 calculation extrapolated to 6.4 kw/ft would result in a PCT of about,1245'F. Therefore, 1320'F determined previously appears to be a good high side estimate of blowdown PCT for a high burnup 15x15 pin.

In both Reference 7 and Reference 2, this blowdown peak occurred between 5 and 9 seconds.

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PAD 3.4 calculations for a 15x15 pin were not perforned in Reference 5, but by i

extrapolating a 17x17 PAD analyses using incremental values from Table 1, it is estimated that the pre-accident 15x15 pin pressure at end of cycle 3 would be about 1500 psi. Based on the pressure decrease calculated for the 15x15 pin in the first 5 seconds in Reference 7, it is estimated that the pin pressure at 5 seconds for a high burnup 15x15 pin would be 1300 psi. The system pressure at that time was determined to be 920 psi. The pressure drop across the clad is therefore 380 psi and the engineering hoop stress is estimated to be 3.0 KPSI.

As shown in Figure 4, this is well below the NUREG-0630 curves and even below the TREAT data. Therefore, it is not expected that any high burnup pins which l

have low LNGRs would experience any early blowdown ruptures.

It should be noted, however, that this is based on extrapolations, and surely direct calculations based on actual condition would be preferable. Also, if indeed high burnups are expected in the future with higher LHGR, this issue i

should be revisited.

In fact, when significant changes in fuel design models and blowdown LOCA models are proposed, this issue should also be addressed.

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7 REFERENCES 1.

R.B.A. Licciardo, "An Evaluation of the Criterion for and the Calculation of Offsite Doses Deriving from Open Containment Purge Valves During a LOCA a Zion Units 1&2," July 20,1989.

2.

N. Zuber, et al., " Quantifying Reactor Safety Margins:

Application of Code $caling, Applicability, and Uncertainty Evaluation Methodology to a Large Break Loss of Coolant Accident " NUREG/CR 5249 (Draft)DraftDate August 1, 1989, 3.

J.M. B' roughton, et al., 'PBF LOCA Test LOC-6, Fuel Behavior Report "

NUREG/CR-3184, April 1983.

4.

D.L. Acey, J.C. Vogelwede, 'A Comparative Analysis of LWR Fuel Designs,"

NUREG-0559, July 1980.

5.

R.A. Weiner, et al., 'laproved Fuel Perforannce Models for Westinghouse Fuel Rod Design and Safety Evaluations," WCAP-10851-P-A, August 1988.

6.

L.J. Stefken

• Developmental Assessment of FRAP-T6" Interim Report No.

EGG-CDAP-5439, May 1981.

7.

L.J. 51efhee; (Personnel communication to G.N. Lauben) " Calculation of Response of Fuel Rod in Zion Reactor During Large Break LOCA " July 18, 1989.

8.

D.A. Powers, R.0. Meyer, " Cladding Swelling and Rupture Models for LOCA Analysis,' NUREG-0630. April 1980.

b Reference 6

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NOTE T0: Frank Miraglia FROM:

Ashok Thadant

SUBJECT:

DPV CONCERNING CONTAINMENT !$0LAT10N VALVES AT ZION 4

The attached provides the response to the question addressed in your note to me dated August 22,1989.

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Enclosure:

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August 25, 1989 h0TE TO: Ashok C. Thadani, Assistant Director f or Systems Division of Engineering 8 Systeins Technology FROM:

Rot'ert C. Jones, Acting Chief Reactor Systems Branch i

Division of Engineering & Systems Technology i

SUBJECT:

DPV CONCERNING CONTAINMENT ISOLATION VALVES AT ZION I

! have reviewed Robert Licciardo's August 22, 1989 memorandum to Frank l

Miraglia wherein he questioned the adequacy of Norm Lauben's analysis transmitted to you by M. W. Hodges note of August 10,1989(subsequently revised by note of August 23,1989). The specific concern was that Norm's analysis failed to consider the consequences of a LOCA using low burnup, low j

pressurefuelandwhichhavehighlinearheatgenerationrates(LNGR)than analyzed by Nom.

t I have further evaluated this issue, with the assistance of Norm Lauben, in response to the memorandum. We have arrived at the following conclusions:

(1) The specific case of low burnup, low pressure fuel is already analyzed in the LOCA analysis. The possibility for fuel pin rupture is continuously examined during the calculation and rupture is not calculated to occur during the early blowdown period of the LOCA.

(2) As burnup increases to 40,000 MWD /MTU, average fuel temperatures i

decreases while fuel pin pressures slowly increase. We judge that this combination of temperatures and pressures would prevent an early rupture during blowdown.

(3) Above 40,000 MWD /MTU. the earlier analysis sufficiently demonstrates that no earlier ruptures occur.

Therefore, the earlier conclusion that early rupture of fuel pins during the blowdown transient is gt credible remains valid, t

i Robert C. J es Acting Chief Reactor Systems Branch Division of Engineering & Systems Technology I

cc: N.Lauben(NLN-353) l 1

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List of Memoranda 1.

Memorandum to T. Murley f rom R. Licciardo dated May 11,1989,subj:

"DifferingProfessionalViewConcerninga)IssuanceofSERtoZion1/2 allowing full power operation with open 42" containment isolation valves, b) Methodology used for calculating related offsite doses."

d 18,1989,subj:

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Memorandum to R. Licciardo from T. Murley(a)ated MayZion 1/2 Containment Isolatii 2.

" Differing Professional View Concerning Valves, and (b) Methodology Used for Calculating Related Offsite Doses."

3.

Memorandum to T. Murley from R. Licciardo dated May 25,1989,subj:

i "Different Professional View (DPV) Concerning Zion. Persons Proposed as

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Third Member of Standing Review Panel."

"DifferingProfessIona.CongelfromT.Murleydated 4.

Memorandum to F. Miraglia, C. E. Rossi F

l View of Robert 8.A. Licciardo May 26, 1989, subj:

Concerning Containment Isolation Valves at Zion."

5.

Memorandum to R. Licciardo from F. Mirag1.in dated June 2, 1989, subj:

" Differing Professional View (DPV) Concerning Containment Isolation Valves at Zion."

6.

Memorandum to R. Licciardo from F. Miraglia dated June 23,1989,subj:

" Differing Professional View (DPV) Concerning Containment Isolation Valves at Zion."

7.

Memorandum to F. Miraglia from R. Licciardo dated June 30,1989,subj:

at Zion."g Professional View (DPV) Concerning Containment Isolation Valves

" Differin 8.

Memorandum to F. Miraglia from R. Licciardo dated July 14,1989,subj:

at Zion."g Professional View (DPV) Concerning Containment Isolation Valves

" Differin l

l 9.

Memorandum to F. Miraglia from R. Licciardo dated July 14,1989,subj:

" Differing Professional View (DPV) Concerning Containment Isolation Valves at Zion." (Note: This memorandum corrected the date given in No. 7.)

I

10. Memorandum to R. Licciardo from F. Miraglia dated July 21,)1989,subj:

"DPV

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Concerning Containment Isolation Valves at Zion (TAC 73427 - Status Information."

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MEMORANDUN FOR: Thomas E. Murley, Director Office of Nuclear Reactor Regulation Robert 8. A. Licciardo, Reactor Engineer (Nuclear)

• FROM:

l Plant Systems tranch Division of Engineering and Systems Technology

SUBJECT:

DIFFERING PROFES$10NAL 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.

ThewritersubmitsaDifferingprofessionalView(DPV)inaccordancewiththe f

provisions of NRC Manual Chapter 4126.

-This issue has arisen out of the Safety Evaluation Report (SER) undertaken for the Zion Units 1 and 2 as prepared by the writerg see Attachment.

The principal issue is the prudent and conservative calculation of the additions to offsite dose which may result from a LOCA at a facility during the use of.

open purge supply and exhaust valves at full power.

The licensee for Zion 1/2 has proposed full power operation of the facility with the 42' purge supply and exhaust containment isolation valves open to alimitedpositionof50'.andcapableofisolationwithinseven(7) seconds of the cessencesent of a LOCA.

The writers SER concludes that the 42' valves at Zion should remain closed in Modes I, 3 3 and 4 because the consequence of the offsite dose to thyroid (from iodine),during a LOCA is unacceptably higha whole body has not been evaluated. The least value for the additional o fsite dose which may be d

proposedwithinthelicensingbasisis64,030renoverthefirstseven(7) seconds of the LOCA. Management staff has disagreed with the writer's methodology and conclusion and plans issuance of a separate SER permitting The writer requests non-issuance of the related SER the operation requested.He also proposes probability of a generic action on other to the licensee.

facilities which have been granted such ifconses based on the staff's current methodology.

In general t'he management staff has adopted a criterion described in SRP BTP C58 64 which is that providing the maximum time for closure of these containment isolation valves does not exceed 5 seconds (and by plant-s acific exception,upto15 seconds),thenthevalveswouldbeclosedbeforetteonset of fuel failure following a LOCA so that the only contribution to offsite dose j

is from ACS operational levels of fission product directly discharged into containment during this period, and then through the open containment isolation valves before closure.

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Ie c-i Thomas E. Murley 2

s/ r In evaluating the consequence for Zion, the writer has used an alternate l

Criterion in BTP CSB 6 4 which-states that:

'The following analyses should be performed to justify the containment purge system design:

An analysis of the radiological consequences of a loss.cf. coolant l

The analysis should be done for a spectrum of break accident.

sites, and the instrumentation and setpoints that will actuate the l

purge valves closed should be identified. The source term used in l'

the radiolgical calculations should be based on a calculation under the terms of Appendix K to detetuine the extent of fuel failure and the concomitant release of fission products, and the fission product activity in the primary coolant. A pre-existing iodine spike should I

be considered in determining primary coolant activity. The volume of containment in which fission products are mixed should be justified, and the fission products from the above sources should be assumed to be released through the open purge valves during the The radiological maximum interval required for valve closure.

consequences should be within 10 CFR Part 100 guideline values.'

UsingtheserelatedguidelinesforZion,(thefuelperformanceoverthe07 secon is detailed and shows that fuel failure byinfringementofDNBRcriteria) occurs within i seconds of the coseencement of the LOCA, and together with other licensing basis responses including fission product release from the fuel gap and the thermal hydraulic conditions in the core, containment and discharge nozzle, result in a substantive discharge of fission products to the environment of far greater consequence than are calculated by the staff.

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 limits, the alternate approach used by the writer shows a substantially increased offsite dose exceeding 10 CFR Part 10011mits, with completely unacceptable consequences to Public Health and Safety.

j The writer requests review of the Differing Professional View in a timely manner in accordance with the provisions of NRC Manual Chapter 4125.

f is4W Robert B. A. Licciardo Registered Professional Engineer California 001056 Nuclear Engineering License No. NU

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Mechanical Engineering License No. M 015380 cc: J. Snier'ek D. Muller S. Varga C. Patel l

F. Miraglia L. Shao A. Thadant J. Werniel J. Eudrick i

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\\ * *..+ h f,f May 11, 1989 Attachment Docket Mos. 50-295 and 50 304 MEMORANDUM FOR: Daniel Muller, Director Project Directorate !!!-2 Division of Reactor Projects !!!, IV, Y and Special Projects FROM:

Jared 5. Wermiel, Acting Chief Plant Systems Branch Division of Engineering and Systems Technology

$USJECT:

0FFSITE RADIOLOGICAL CON 5(QUENCES OF LOCA DURING CONTAINMENT PURGE PROPOSED IN T5 CHANGES FOR ZION 1 AND 2

Reference:

LettertoN.R.Denton(NRC)FromP.C.Leonarddated February 2,1986,

Subject:

Zion Nuclear Power Station, Units 1 and 2 Proposed Amendannt to Facility Operating License No. DPR-3g and DPR-48' Plant Name:

Zion Nuclear Power Station, Units 1 and 2 Licensee Cosmonwealth Edison Company TAC Nos.:

55417 and 55418 Review Status:

Complete ZionUnits1and2(Ceco)hasrespondedtoanNRCrequesttoproposeT5to priserily constrain operation of the large (42") containment purge supply and exhaust valves on these units see reference 1.

The former Plant Systems Branch, Section A, of the Division of PWR Licensinp A requested Section B of the same branch to review the offsite radfologica consequences of this proposal.

The enclosed Safety Evaluation Report has been prepared by the technical reviewer initially assigned to this task, namely Robert 5. A. Licciardo.

The licensee's pro sal is to allow full power operation of the facility with and exhaust containment isolation valves open to a the 42' purge supp limited position o 50', and capable of isolation within seven (7) seconds of the commencement of a LOCA.

The review concludes that the 42' valves at Zion should remain closed in Modes 1, 2, ) and 4 because the consequence of the offsite dose to thyroid (fromiodine).duringaLOCAisunacceptablehightwholebodydosehasnotbeen The least value for the additional offsite dose which any be proposed evaluated:

withinthelicensingbasisis64,000removerthefirstseven(7) seconds.

The conventional treatment of BTP CSB 6-4 which assumes that fuel failure does not occur over the first 515 seconds after a LOCA and thereby that only RCS and enerating inventory of fission products is released to the containmen analyses for containment response, and Itcensing basis requirements (including criteria)forthecalculationfor,andtheoccurrenceof,fueldamageandthe

- *a+d*4a+4aa and enatsent of resultine source terms.

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s? r Our $ ALP input is provided in Enclosure 2.

We consider our efforts on, TAC Nos. 55417 and 56418 to be complete.

Jared 5. Wermiel, Acting Chief Plant Systems Branch Division of Engineering and Systems Technology

Enclosures:

As stated cc w/ enclosures C. Pate)

CONTACT: R. Licciardo X20076 t

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Jared S. Wermiel, Acting Chief Plant Systems Branch Division of Engineering and Systems Technology

Enclosures:

As stated cc w/ enclosures:

C. Patel-CONTACT: R. Licciardo-120876 a

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PLANT SYSTEMS BRANCH 0FF$!TE RADIOLOGICAL CONSEQUENCE OF LOCA DURING CONTAINNENT PURGE ZION NUCLEAR POWER STATION, UNITS 1 AND 2 D0CKET N05. 50 295 and 50-304 t

=1.0 !NTRODUCTION Zion Units 1 and 2 (Ceco) has responded to an NRC request to propose TS to primertly constrain operation of the large (42") containment purge supply and exhaust valves on these units.

p The former Plant Systems Branch, Section A, of the Division of PWR Licensink I

A, requested Section B.of the same branch to r'eview the offsite radiologice consequences of this proposal.

L

.2.0 EVALUATION Background review shows that the facility was evaluated on the basis of normally closed purge valves so that these consequences were never included Further,thataletterfromWestinghouseJW).toCosmonwealth l'

I in the Zion SER.

. Edison Company dated October 22 1976onthesubjectof*0fWiteDosesDuring l

- LOCA and Containment Purge" (Ref. 2) has never been evaluated by the NRC.

l subsequent to the TMI-2 event, the operability ano automatic control of these valves was evaluated leading to the request for the required T t

3)whichwas intended to be resolved in a subsequent probabilistic risk assessment which definitivelyexcludeditfromconsiderationwithoutany; justification (Ref.4).

l uses an RCS The W analyses undertaken under Commonwealth Edison instruction,f the accident i

l operational inventory of 60 oc/gm equivalent I 131 at the time o with a resulttog site boundary thyroid dose due to iodin 6 (during closure of the valves).- of 52 rom, and which added to the containment leakage dose of 123

. rem gives a total 175 rem which is within the 10 CFR 100 limit of 300 rem.

The total iodine inventory of the RCS is assumed to be released into containment on initiation of the LOCA; a 505 plate out is assumed leaving the residual 505

as part of containment inventory for discharge out through both fully open

)-

containmentpurgelinesforatotalofseven-(7 seconds).

However, whed reviewed against the BTP CSB 6-4, Item B.S.a requires that:

"The source tem 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 concosmitment release of fission products, and the fission product activity in the primary coolant."

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i,/ r SRP 4.2 identifies fuel failure with infringement of DNBR crJteria, Further:

with the related requirement that gap activity be considered as part of the source ters, and Regulatory Guide 1,77 reconnends that under similar

^

Fuel circumstances, gap activity should be assumed at 10% of core activity.

damage criteria also includes the occurrence of center line telting w but the Zion SAR shows this does not occur.

RevisingthesourceterstoAppendixKcalculations[inwhichallfuelgoes toDNBRinisecond)withrelatedreleaseofallgapactivityintocontainment, with limited blowdown to offsite during the related 7 seconds closure time and absent a 505 plate out of iodine as can be interpreted from the above referenced item 8.5.a. increases offsite dose due to containment purge above rem and would thereby be completely unacceptable.

by a factor of 3400 to 176 000Limitingthepurgelinevalvestoanopeningof50'couldre to 64,000 rem and represents the least value which may be proposed within the licensing basis.

The BTP CSB 6-4 proposing that valve closure within 5 seconds will Note:

ensure purge valves are closed before the onset of fuel failures has since

Further, been extended by the staff on a plant-specific basis to 15 seconds.

the writer cannot find any safety evaluation report supporting these positions.

ThesepositionscannotbesustainedforZionsincea)DNBRinfringement(from Appendix K calculations) and hence fuel failure and gap activity release [Ref.

SRP 4.2) of 105 of core inventory (Ref. Regulatory Guide 1.77) occur within i second of the initiation of the LOCA, b) related maximum clad temperatures of 1750'F occur ismediately and never reduce below 1400*F, c) RCS pressure in the region of the core rapidly reduces from 2250 psia to g00 psia in 7 seconds increasing potential pressure drop across the cladding for release of gap activity to the RCS inventory, d) the massive bulk boiling and blowdown 270,000 lbs of RCS inventory surrounding the failed fuel ultimately discharges 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 2x42' fully open lines, or 5400 lbs for the same lines with valve closed to 50'.

3.0 CONCLUSION

2, 3 d 4 because The 42' valves at Zion should remain closed in Modes 1the consequen b

during a LOCA The least value is unacceptably high; whole body dose has not been evaluated.

for offsite dose to the thyroid which may be proposed within the existing licensing basis is 64,000 rem.

The conventiopal treatment of BTP CSB 6-4 which assumes that fuel failure does not occur over the first 5-15 seconds after a LOCA and thereby that only RCS operating inventory of fission products is released to the containment, and then to the environment, cannot in general be sustained against thermal hy(draulic including analyses for containment response, and licensing basis requireme quantification and treatment of the resulting source terms.

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References o

i LetterfromP.C. Blond (Ceco)toH.R.Denton(NRC);

Subject:

Zion, Units 1 and 2, Proposed Amendment to Facility Operating License Nos. DPR-39 and DPR-48 dated February 21, 1986.

t Letter from R. L. Kelley LW) to C. Reed (Ceco);

Subject:

Offsite 2.

Oose During LOCA and Containment Purge, dated October 22,1986.

i Letter to L. 0. De1 George (Ceco) from S. A. Varga (NRC);

Subject:

3.

Generic Concerns of Purging and Venting Containments, dated l'

September 9, 1981.

l-Memo for F. H. Robinson from R. W. Houston,

Subject:

" Evaluation 4.

of the Risk at Zion," dated August 14, 1985.

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SPLB SALP INPUT Units 1 and 2 Zion Nuclear Generating Stations, Operation Plant Name:

Containment Purge and Vent Valve SER

Subject:

TAC Nos.:

55417/8 Summary of Review / inspection Activities.

The licensee provided an evaluation of offsite doses undertaken in 1976. This was undertaken with a methodology and source term chosen by the licensee. The licensee did not present results from alternative more detailed methodologies which could be considered enforceable under existing regulatory positions and the related circumstances.

Narrative Discussion of Licensee Performance - Functional Area The single.only methodology used by the licensee is not an acceptable' approach' for estimating doses under the proposed circumstances and especia1Ty since alternate detailed evaluations required by the SRP give greatly increased values beyond 10 CFR Part 100 limits. A prudent approach would have recgnized the deficiencies and risks in the single methodology adopted with resulting substantively different recommendations to ensure public health and safety.

Author: Robert B. A. Licciardo Date:

May !!, 198g e

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.a.uwovow, o. c. mu ItAY 181989 MEMORANDUM FOR: Robert Licciardo, Reactor t

Engineer (Nuclear)

Plant Systems Branch Division of Engineering and Systems "echnology FRON:

Thomas E. Murley, Director Office of Nuclear Reactor Regulation

-$UBJECT:

DIFFERINGPROFES$10NALVIEWCONCERNING(A) ZION 1/2 CONTAINMENT!$0LAT!0NVALVES,AND(B)METHODOLOGYUSED FOR CALCULATING RELATED OFFSITE D0SES o

l l

This is to acknowledge that on May 12,1989 I received your Differing i -

ProfessionalView(DPV)concerningthecaptionedsubject. Please submit a listing of persons you would like me to consider as the third member of the-Standing Review Panel and as an alternate mestier for the Standing Review Panel.

The standing Review Panel will determine within 7 days if adequate information has been supplied to initiate a review of your q'V.

. n o n

L is E Murl ter ffice of Nuc1 Ru or Regulation i

cc: J. Snierek L

F. Mirap11a J. Part ow r

5. Varga

8. Holahan E. Rossi L. Shao C. Patel CONTACT:

H. Smith, PMA5 121287 y

.qqqsa $94i ' Q-

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NUCLEAR REGULATORY COMMISSION WA&MINGTON, D. C. 20666 (e...*

E 2 8 1999 MEMORANDUM FOR:

Thomas E. Murley, Director Office of Nuclear Reactor Regulation FROM:

Robert B. A. Licciardo Plant Systems Branch Division of Engineering and Systems Technology

SUBJECT:

DIFFERENT PROFESSIONAL VIEW (DPV) CONCERNING ZION.

PERSONS PROPOSED AS THIRD MEMBER OF STANDING REVIEW PANEL.

- On May 19 I received your request to submit a listing of persons-to consider as the third (and) alternate member of the Standing Review Panel for the purpose of reviewing _the writers D.P.V dated May 11, 1989.

For this purpose I l

nominate:

i Steven A. Varga, Director, Division of Reactor Projects Gary M. Holahan, Acting Associate Director for Regions III and V Frank J. Congel. Director, Division of Radiation Protection and Emergency Preparedness My understanding from NRC Appendix 4125 Section 8.1 is, that the current role ofthepanelistodetermineifenoughInformationhasbeensuppliedto undertake a detailed review of the issue.

And that given a favorable review, l

the necessary interdisciplinary expertise can be assembled to formulate a final disposition. On this basis, the above persons are nominated.

M l

Robert B. A. Licciardo

(_

Registered Professional Engineer, California Nuclear Engineering License No. NU001056 Mechanical Engineering License No. M015380 cc:

J. Sniezek F. Miraglia' J. Partlow o-j S. Varga G..Holahan E. Rossi L.-Shao l

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f wasmNoTow. c. c. Posts NAY 2 6 1989 MEMORANDUM FOR:

Frank J. Miraglia, Associate Director for Inspection and Technical Assessment, NRR C.-Ernie Rossi, Director Division of Operational Events Assessment NRR Frank J. Congel, Direetor Division of Radiation Protection and Emergency Preparedness, NRR.

FROM:

Thomas E. Murley, Director Office of Nuclear Reactor Regulation

SUBJECT:

DIFFERING PROFES$10NAL VIEW 0F ROBERT B.A.-LICCIAR00 CONCERNING CONTAINMENT ISOLATION VALVES AT ZION

\\

Enclosed is a memorandum from Mr. Licciardo to Dr. Murley, dated May 11, 1989 expressing a Differing Professional View.

In accordance with NRC Manual Chapter 4125 and NRR Office Letter No. 300 dated March 24, 1989, you are hereby designated as-the Panel to review and recomunend to the Director,. NRR the appropriate disposition of Mr. Licciardo's Differing Professional View.

If you deem it necessary, you may solicit input fmm other NRR technical staff or contractors.

In carrying out your review and formulating your recommendations to me, you.

should be guided by the Appendix'to NRC Manual Chapter 4125 with special emphasis on Sections B.6 and 8.7.

)

Thomas E. Murley, uirector I

Office of Nuclear Reactor Regulation

Enclosure:

As stated cc:

J. H. Sniezek l

J. Larkins R..Licciardo

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l MEMORANDUM FOR: Thomas E. Murley, Director Office of Nuclear _ Reactor Regulation Robert 8.A.Licciardo,ReactorEngineer(Nuclear)-

FRON:

Plant Systems Branch Division of Engineering and Systems Technology L

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.

The writer submits a Differing Professional View (DPV) in accordance with the provisions of NRC Manual Chapter 4125.

'i This issue has arisen out of the Safety Evaluation Report (SER) undertaken for the Zion Units 1 and 2 as prepared by the writer; see Attachment.

The principal issue is the prudent and conservative calculation of the additions to offsite dose which may result from a LOCA at a facility during the use of open purge supply and exhaust valves at full power.

- The licensee for Zion 1/2 has proposed full power operation of the facility with the 42' purge supply and exhaust containment isolation valves open to alimitedpositionof50',andcapableofisolationwithinseven(7) seconds of the commencement of a LOCA.

The writers SER concludes that the 42' valves at Zion should remain closed 3 and 4 because the consequence of the offsite dose to thyroid in Modes 1, 2,during a LOCA is unacceptably high whole body has not been (fromfodine)Theleastvaluefortheadditionaloffsitedosewhichmaybe evaluated.

proposedwithinthelicensingbasisis64,000renoverthefirstseven(7) seconds of the LOCA. Management staff has disagreed with the writer's methodology and conclusion and plans issuance of a separate SER permitting The writer requests non-issuance of the related SER the operation requested.He also proposes probability of a generic action on other to the licensee.

facilities which have been granted such licenses based on the staff's current methodology.

In general, the management staff has adopted a criterion described in SRP BTP CSB 6-4 Which is that providing the maximum time for closure of these containment isolation valves does not exceed 5 seconds (and by plant-sweific exception, up to 15 seconds), then the valves would be closed before tse onset of fuel failure following a LOCA so that the only contribution to offsite dose is from RCS operational levels of fission product directly discharged into containment during this period, and then through the open containment isolation

-valves before closure.

S 9 f? L ? h.L.L.

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y Thomas E. Murley '

s.1 r In evaluating the consequence for Zion, the writer has used an alternate Criterion in BTP CSB 6-4 which states that:

'The following analyses should be performed to justify the containment purge system design:

An analysis of the radiological consequences of a loss-of-coolant accident. The analysis should be done for a spectrum of break sizes, and the instrumentation and setpoints that will actuate the purge valves closed should be identified. The source term used in the radiological calculations shou'1d be based on a calculation under the terms of Appendix K to determine the extent of fuel failure and the concomitant release of fission products, and the fission prvduct activity in the primary coolant. A pre-existing iodine spike should i

be considered in determining primary coolant activity. T w volume of containment in which fission products are mixed should be L

l justified, and the fission products from the above sources should be assumed to be released through the open purge valves during the maximum interval required for valve closure. The radiological L

consequences should be within 10 CFR Part'100 guideline values."

Using these related guidelines for Zion,(the fuel performance over the 0-7 secon l

is detailed and shows that fuel failure byinfringementofDNBRcriteria) occurs within i seconds of the connencement of the LOCA, and together with other licensing basis responses including fission product release from the fuel gap

(

and the thermal hydraulic conditions in the core containment and discharge nozzle, result in a substantive discharge of fission products to the environment of far greater consequence than are calculated by the staff.

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 limits, the alternate approach used by the writer shows a substantially increased offsite dose exceeding 10 CFR Part 100 limits, with completely unacceptable consequences to Public Health and Safety.

The writer requests review of the Differing Professional View in a timely manner in accordance with the provisions of NRC Manual Chapter 4125.

f 654W Robert B. A. Licciardo Registered Professional Engineer California Nuclear Engineering License No, NU 001056 L

Mechanical Engineering License No. M 015380

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J. Snier'ek D. Muller S. Varga C. Patel F. Miraglia L. Shao A. Thadani J. Wermiel J. Kudrick

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Attachment Docket Nos. 50-295 and 50-304 -

MEMORANDUM FOR: Daniel Muller, Director Project Directorate !!!-2 Division of Reactor Projects !!!, IV, Y and Special Projects FRON:-

Jared 5. Wermiel, Acting Chief Plant Systems Branch Division of Engineering and Systems Technology L

SU5 JECT:

0FF51TE RADIOLOGICAL CONSEQUENCES OF LOCA DURING l

CONTAllg1ENT PURGE PROPOSED IN T5 CHANGES FOR ZION 1 AND 2-

Reference:

LettertoH.R.Denton(NRC):FromP.C.Leonarddated-February 2,1986,

Subject:

Zion Nuclear Power Station, Units 1 and 2 Proposed Amendment to Facility _ Operating License No. DPR-39 and DPR Plant Name:

Zion Nuclear Power Station, Units 1 and 2 Licensee:

Cossenwealth Edison Comparty TAC Nos.:.

55417 and 55418 Review Status:

Complete IfonUnits-1and2(Ceco)hasrespondedtoanNRCrequesttoproposeTSto primarily constrain operation of the large (42") containment purge supply and exhaust valves on these units; see reference 1.

The former Plant Systems Branch, Section A. of the Division of PWR Licensing I

A, requested Section B of the swee branch to review the offsite radiological

consequences of this proposal.

The enclosed Safety Evaluation Report has been p-epared by the technical reviewer initially assigned to this task, namely Robert B. A. Licciardo.

The licensee's proposal is to allow full power operation of the facility with J

the 42" purge supply and exhaust containment isolation valves op(en to a7)secondsof limited position of 50*, and capable of isolation within seven

'the emmencement of a LOCA.

'The review concludes that the 42' valves at Zion should remain closed in Modes 1,2,13. and'4 because the censequence of the offsite dose to thyroid

-(from todine) during a LOCA is unacceptable high; whole body dose has not been The least value for the additional offsite dose which may be proposed evaluated:

within the licensing basis is 64,000 rem over the first seven (7) seconds.

The conventional 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 that only RCS operating inventory of fission products is released to the containment, and Ltheir to the environment, cc! mot in general be sustained against thermal hydraulic o

= analyses for containment response, and licensing basis requirements (including critpla)forthecalculationfor,andtheoccurrenceof,fueldamageandthe qusatification and treatment of resulting source terms.

pg

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Daniel Muller '

sf /*

Our SALP.in'put is provided in Enclosure 2.

We consider our efforts og TAC Nos. 55417 and 55418 to be complete.

Jared S. Wermiel, Acting Chief Plant Systems Branch Division of Engineering and Systems Technology

Enclosures:

As stated cc w/ enclosures C. Pate)

-CONTACT: R. Licciardo

^

X20876 4

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- Daniel Muller l

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• i-Dur SALP input is provided in Enclosure 2.

We consider our efforts os TAC Nos. 55417 and 55418 to be complete.

Jared S. Wermiel, Acting Chief Plant Systems Branch Division of Engineering and Systems Technology

Enclosures:

As stated cc w/ enclosures:

C. Patel CDNTACT: R. Licciardo

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Eulosu$1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PLANT SYSTEMS BRANCH 0FFSITE RADIOLOGICAL CONSEQUENCE OF LOCA DURING CONTAINMENT PURGE ZION NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET N05. 50-295 and 50-304

1.0 INTRODUCTION

l ZionUnits1and2(Ceco)hasrespondedtoanNRCrequesttoproposeTSto-primarily constrain operation of the large (42") containment purge supply

.and exhaust valves on these units.

The former Plant Systems Branch, Section A, of the Division of PWR Licensing A, requested Section 8 of the same branch to' review the offsite radiological consequences of this proposal.

240 EVALUATION Background review shows that the facility was evaluated on the basis of normally closed purge valves so that these consequences were never included in the Zion SER. Further, that a letter from WestinghouseiW) to Cosmonwealth LOCA and Containment Purge" (Ref.1976 on the subject.cf '0frsite Doses D

-Edison Company dated October 22 2)hasneverbeenevaluatedbytheNRC.

Subsequent to the TNI-2. event, the operability and automatic control of these valves was evaluated leading to the request for the required TSRadiolog 3)whichwas l

intended to be resolved in a subsequent probabilistic-risk assessment which l

definitively excluded it from consideration without ary justification (Ref. 4).

uses an RCS The W analyses undertaken under Cosmonwealth Edison instruction,f the accident r

operitional inventory of 60 oc/gm equivalent-! 131 at the time o with a resulting site boundary thyroid dose due to iodine (during closure of l

the valves), of 52 ren, and which added to the containment leakage dose of 123 rem gives a total 175 rem which is within the 10 CFR 100 limit of 300 rem.

The total iodine inventory of the RCS is assumed to be released into containment on initiation of the LOCA; a 505 plate out is assumed leaving the residual 505 as part of containment inventory for discharge out through both fully open containment purge lines for a total of seven (7 seconds).

However, when' reviewed against the BTP CS8 6-4, Item B.5.a requires that:

"The source 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 concosuitment release of fission ptmducts, and the fission product activity in the primary coolant."

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Further: SRP 4.2 identifies fuel failure with infringement of DNBR criteria, with the related recuirement that gap activity be considered as part of the source term, anc Regulatory Guide 1.77 reconnends that under similar circumstances, gap activity should be assumed at 105 of core activity. Fuel damage criteria also includes the occurrence of center line melting'with measures of additional activity release also guided by Regulatory Guide 1.77, but the Zion SAR shows this does not occur..

Revising the source ters to Appendix K calculations (in which all fuel goes to DNBR in i second) with related release of all gap activity into containment, with limited blowdown to offsite during the related 7 seconds closure time and absent a 50% plate out of iodine as can be interpreted from the above referenced item B.5.a. increases offsite dose due to containment purge above L

by a factor of 3400 to 176 000 rem and would thereby be completely unacceptable.

i Limitingthepurgelinevalvestoanopeningof50'couldreduceoffsitedose to 64,000 rem and represents the least value which may be proposed within the L

licensing basis.

Note: The BTP CSB 6-4 proposing that valve closure within 5 seconds will ensure purge valves are closed before the onset of fuel failures has since been e nended by the staff on a plant-specific basis to 15 seconds. Further.

the writer cannot find any safety evaluation report supporting these positions.

These positions cannat be sustained for Zion since a) DNBR infringement (from Appendix' K calculations)-and hence fuel failure and gap activity release [Ref.

SRP 4.2) of 105 of core inventory (Ref. Regulatory Guide 1.77) occur within i second of the initiation of the LOCA, b) related maximum clad temperatures of 1750*F occur immediately.and never reduce below 1400'F. c) k!! pressure in the reg'in of the core rapidly reduces from 2250 psia to g00 psia in 7 seconds increasing potential pressure dro f gap activity to the RCS inventory, d)p across the cladding for release o the massive bulk boiling and blowdown surrounding the failed fuel ultimately discharges 270,000 lbs of RCS inventory 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 5400 lbs for the same lines with valve closed to 50'.

3.0 CONCLUSTON 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) durin a LOCA is unacceptably hight whole body dose has not been evaluated. The east value for offsite dose to the thyroid which may be proposed within the existing licensing basis is 64,000 ren.

The conventic'nal treatment of BTP CSB 6-4 which assumes that fuel failure does not occur over the first 5-15 seconds after a LOCA and thereby that only RCS operating inventory of fission products is released to the containment, and then to the environment, cannot in general be sustained against thermal hydraulic analyses for containment response, and licensing basis requirements (including criteria)forthecalculationfor,andtheoccurrenceof,fueldamageandthe quantification and treatment of the resulting source terms.

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f-Reference _s o

1..

LetterfromP.C. Blond (Ceco)toH.R.Denton(NRC);

Subject:

Zion, Units 1 and 2, Proposed Amendment to Facility Operating bicense Nos. DPR-39 and DPR-48 dated February 21, 1986.

2.

Letter from R. L. Kelley [W) to C. Reed (Ceco);

Subject:

Offsite Dose During LOCA and Cont 7a nment Purge, dated October 22, 1986.

3.-

Letter to L. O. De1 George (CECO) from S.A. Yarga (NRC);

Subject:

Generic Concerns of Purging and Venting Containments, dated September 9, 1981.

4.

Memo for F. H. Robinson from R. W. Houston,

Subject:

" Evaluation of the Risk at Zion,' dated August 14, 1985.

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.! r SPLB SALP INPUT Plant Name:

Zion Nuclear Generating Stations, Units 1 and 2 SER-

Subject:

- Containment Purge and Vent Valve Operation TAC Nos.:

55417/8

$unenary of Review / inspection Activities The licensee provided an evaluation of offsite doses undertaken in 1976. This was undertaken with-a methodology and source term chosen by the licensee. The licensee did not present results from alternative more detailed methodologies which could be considered enforceable under existing regulatory positions and the related c,1rcumstances.

Narrative Discussion of Licensee Performance - Functional Area The single only methodology used by the licensee is not an acceptable approach

-for estimating doses under the proposed circumstances and especially since alternate detailed evaluations required by the SRP give greatly increased o

values beyond 10 CFR Part 100 limits. A prudent approach would have recoplnited the deficiencies and risks in the single methodology adopted.with resu ting substantively different recossendations to ensure public health and l

safety.

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Author: Robert 8. A. Licciardo

-Date:

May 11, 1989 i

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Docket Nos. 50-295 and 50-304 MEMORANDUM FOR: Daniel Muller, Director Project Directorate III-2.

Division of Reactor Projects !!!, IV, Y andSpecial' Projects FROM:

Jared S. Wemiel Acting Chief l5 Plant Systems Branch

. Division of Engineering and Systems Technology I

$UBJECT:-

PROPOSED TS CHANGES ON PURGE / VENT OPERATION

Reference:

~

Zion Nuclear Power Station Units 1 and 2 Proposed Amendment to FacilityL0perating~ License No. DPR-3g and DPR-48, letter to H. R. Denton (NRC) From P. C. Leonard dated February 2,1986 Plant Name:

Zion Nuclear Power Station Units 1 and 2

. Licensee:

Cosmonwealth Edison Company' Revtow Status:. -Complete The' Plant Systems Branch has reviewed Cosmonwealth Edison's proposed ' changes to l

-the Technical Specifications on containment purge and vent valve operation for l

Zion Units 1 and 2.-as described in:a letter dated February 21,1986. The-proposed changes are either administrative in nature or are to comply with the generic concerns of MPA B-24 as:it is related to demonstration of containment purge and-vent valve. operability. Based on the enclosed safety evaluation report (Enclosure 1), the Plant Systems Branch concludes that the proposed Technical Specifications are acceptable.

There-is one possible follow-up item that should be clarified with the Itcensee,however. There is some question as to how the licensee intends to-

preclude opening the~ purge / vent valve beyond the 50 degree angle as specified in the TS.- DiscussionswiththeMechanicalEngineeringBranch(MEB)have'

. indicated that a positive stop is required on the valve to prevent opening i beyond the TS angle. Operational procedures, by themselves, are not acceptable.

Since none of the incoming information addresses how the opening will be limited,:the' Project Manager should verify with the licensee that a positive E

stop has been installed on the valve. If this is not the case, this issue g

should be pursued with ME8.

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i Dur SALP' input is provided in Enclosure 2.

We consider our efforts on TAC

Nos. 55417 and 55418 to be conplete aA. cJ12

, red S. Wermiel, Acting Chief Plant Systems 8 ranch Division of Engineering and Systems Technology

Enclosures:

As stated cc w/ enclosures:

C. Patel C0hTACT: J. Kudrick

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e SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PLANT SYSTEMS. BRANCH PROPOSED TECHNICAL SPECIFICATIONS r

CONTAINMENT PURGE IION NUCLEAR POWER STATION, L' NITS 1-AND 2 o

DOCKET N05. 50-4g5 and 50-304

2.0 INTRODUCTION

Cosmonwealth Edison, the owner of the nuclear power plants Zion Units 1 and F, 1986, an amendment to p*oposed in a letter to H. Denton dated February 21,The amendment proposed Facility Operating License Nos. DPR-39 and DPR-48.

changestotheTechnicalSpecifications(TS)relatedtoventandpurge i

o wrations as well as restricting the maximum purge valve position. These cian es were in response to an NRC request in a, Safety Evaluation Report dated Apri 3, 1984 Simply stated, the request wat to reflect the permissible owration of the purge and vent valves into the TS. The submittal contains tie requested changes.

'2.0 EVALUATION The proposed changes related to restrictions in purge and vent operations.

Specifically, they include the allowable angle the purge supply and exhaust valves can be opened, the number of valves that can be used at one time, the valve closure time, and the goal for purging time in one year. Each of these changes will be discussed below.

However, before the individual T$ changes.are discussed, there is one t

survie11ance test that was recossended in the staff SER that was not added to the proposed TS. The staff had recommended the periodic leakage testing of the valves with resilient seals. The frequency was to be once per three months during operating Modes 1~through 4. if the valves were considered to-be-active.

In response'to this request, the licensee indicated that the additional

= surveillance requirement was not needed for the valves at Zion because the isolation valve seal water system and penetration pressurization system are designed to continuously detect any leakage during plant operation. If leakage is detected, an alarm is sounded in the control room. The staff has reviewed the -licensee's justification for not performing the added leakage tests. As pact of their justification, the licensee, in the bases Section 3.4 of the TS, indicated that the seal water is introduced at a pressure of 50 psig. This pressure is slightly higher than the peak containment post accident pressure. Further, the seal water system and penetration pressurization system are included in TS Section 3.g.1 and 3.g.2 which includes limiting condition for operation (LC0) and surveillance requirements.

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Based on the above, the staff concludes that the' continuous leakage detection systems now in place at Zion Units 1 and 2 satisfy the requirements of the In addition, the surveillance leakape tests referenced in the staff's SER.

current TS on the eskape systems meets the intended purpose of the. suggested added TS..Therefore, the staff concurs with the licensee that no additional surveillance testing or added TSs are necessary.

The proposed TS indicating that the purge supply and exhaust valves shall not be opened more than 50 degrees is consistent with the staff's SER dated April 3, 1984 Therefore the staff finds the proposed TS acceptable. The acceptance of the allowable opening angle is based in part, on the demonstration of acceptable stresses within the valve. Inequallyimportantparameterin determining the closure stresses is the closure time. The staff concluded, as documented in the April,1984 SER, that acceptable closure times range between 5 and 8 seconds. The proposed TS change, in this regard, is to change the survie11ance test value from the current 60 seconds to 7 seconds. The revised closure time reflects the acceptable stress analysis and is therefore acceptable.

Another proposed change is to assure that the containment purge valves shall not be open concurrently with the containment vent valves. This operational restriction is consistent with the guidelines set forth in SRP Section 6.2.4 to minimize the number of pathways open at any one time. Based on this compliance with the SRP, the staff finds the operational guidance provided for vent and purge operation acceptable.

An important consideration in the development of an effective program is the selection of a usage factor as well as the reasons for vent and purge The licensee has proposed a goal of 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year. This operation.

time has been established based upon the licensee's estimate to-limit the concentration of radioactive materials in the containment atmosphere to less than 100 times the maximum _ permissible concentration.per 10 CFR 20. After 4-review of the purging criteria, the staff has concluded that the program including the goal. established by the licensee is acceptable. However, due to the importance the staff has placed on the need to minimize purging or venting of the containment, the staff believes that additional clarification should be added to the T5 to ensure that purging be performed only for safety related A marked up copy of the appropriate TS page is enclosed which the reasons.

staff would find acceptable. The licensee has agreed to the staff's proposed markup in a series of telephone conferences. Based on'the verbal agreement of the~ marked up changes, the staff finds the proposed use of the purge and vent systems acceptable..

An additional consideration must be included in the overall evaluation of the purging program, in light of the fact that large diameter valves are being For these conditions, SRP Section 6.2.4, periods greater than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br />. indicates that the radiological conse used for time The with the purge / vent valves assumed open at time zero must be calculated.

analysis should show that 10 CFR part 100 limits are not exceeded.

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Guidance is provided in the SRP concerning the source term to be used for the dose consequences due to the release through the valves until calculating'he guide indicates that for valve closure times within five closure.

-interpreted by the staff to mean that only the pre-existing iodine s,has been seconds, isolation is assured prior to onset of fuel failure. This pike need c

to be considered in deternining primary coolant activity without thi need for L

further justification. Fcr closure times slightly beyond 5 seconds, the staff has evaluated the merits of assuming no fuel failure on a case by case basis.

L Consideration has included the transport times necessary to sweep the source from the failed fuel into the reactor coolant, from the fuel pins to the postulated pipe rupture, fron the pipe rupture to the nearest pipe inlet of the open purge line, and filally through the duct to the isolation valve.

L Based on this ratiorale, the staff has concluded that there will be a substantial time delay between the onset of fuel failure and the actual release of products from the containment as a result of the fuel failure.

Additionally, there will be a finite minimum time before initiation of fuel failure can occur. Using the above rationale, the staff has concluded that's more reasonable upper bound of valve closure time for which no source term contribution due to fuel failure can be conservatively assumed is 15 seconds.

Therefore, for the Zion closure time of seven seconds, the staff has concluded that fuel failure need not be considered. Based on the above, the staff has concluded that only the pre-existing iodine spike need be considered.

The licensee has computed the dose consequences considering the above source ters.. The results show that using a 60 uc/gm equivalent 1-131 spike at the time of the accident, the sitt' boundary thyroid dose due to iodine up until valve closure is 52 rem. When ndded to the containment leakage dose of 123 rem yields a total dose of 175 rem. This is well within 10 CFR 100 requirements of 300 rem.

The staff has performed an independent calculation of the dose contribution due to releases through the purge / vent pathways. The results confirm the-licensee's value. Based on this agreement, the staff finds that the dose consequences due to purging operations are acceptable and within 10 CFR 100 limits.

3.0 CONCLUSION

i Based on the above evaluation, the staff concludes that the preposed changes-to the Zion Units 1 and 2 Technical Specifications for limitation on purge and vent valve operation above cold shutdown are more restrictive than current TSs and consistent with the commitments identified in the staff SER on the same subject. Therefore, the staff finds the proposed changes acceptable.

5520 NAME: I'lon TACS 55417/8

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SPLB SALP 1NPUT d

. Plant'Name:

Zion Nuclear Generation Stations. Units 1 and 2 SER

Subject:

Containment Purge and-Vent Yalve Operation TAC Nos.:.

55417/8 Sver.ary of Review / inspection Activities i

The licensee initially proposed Technical specification changes for

However, containment purge and vent valve operation needed revision.

data revisions adequately addressed the concerns.

Narrative Discussion of' Licensee Performance - Functional Area The licensee's approach for resolution of generic concerns related to the demonstration of containment purge and vent valve was viable and sound from a safety standpoint.

' Authors:

J. Kudrick and C. L1 Date:; May 10,1989 f

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June 2,1989 J

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MEMORAkDUM FOR: Robert Licciardo Reactor Engineer L

Plant Systems Branch Division of Engineering and Systems Technology FROM:-

Frank J. Miraglia, Associate Director i

for Inspection and Technical Assessment i

SUBJECT:

DIFFERINGPROFESSIONALVIEW(DPV)CONCERNINGCONTAINMENT ISOLATION VALVES AT ZION In accordance with NRC Manual Chapter 4125 and NRR Office Letter 300 the Standing Review Panel of Frank Miraglia, Charles E. Rossi and Frank d,ongol reviewed the material submitted to Dr. Murley on the subject matter. The Panel has determined that adequate information has been supplied to initiate a review of your DPV.

It is our intent to meet with you in the near future.

prtstant es p *4 W 188 presh 3. F1888 Frank J. Miran11a, Associate Director I

for InspectLon and Technical Assessment ec:

J. Snierek J. Larkins C. E. Rosst F. Congel DJITR3UTION KDT R/F FMiraglia HSmith g gee $i; d4 m. m 2..

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i MEMORANDUM FOR: Robert Licciardo Reactor Engineer L

Plant System 1 Branch Division of Engineering and

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Systers Technology FROM:

Frank J. Miraglia, Associate Director for intpection and Technical Assessment

SUBJECT:

DIFFERINGPROFESSIONALVIEW(DPV)CONCERNINGCONTAINHENT ISOLATION VALVES AT ZION 4

The Standing Review Panel of Frank Miraglia, Charles E. Rossi and Frank Congel reviewed the material submitted to Dr. Murley on the subject matter. The Panel met with you on Friday, June 16, 1989 to further discuss your views. At that meeting the Panel requested that you more clearly state-your concern regarding the time to fuel failure used in LOCA analyses. The Panel also requested that you also clarify the mechanisms for transporting fission products from the

. primary to containment used in your analyses.

In addition, the Panel requested that you provide your view as to the safety significance of-proceeding with the proposed Zion amendment and the safety significance of your concern regarding LOCA analyses.

I Please let me knw when you will provide the requesud infomation. As we have indicated to you previously it is our intent to comply with the milestones in NRC Manual Chapter 4125 and NRR Office Letter 300.

Oristmal signed by henkI.Suesue Frank J. Miraglia, Associate Director for Inspection and Technical-Assessment cc:

J. Sniezek J. Larkins C. E. Rossi F. Congel DISTRIBUTION Central File ADT/RF 4FMiraglia e HSmith DIFFERING PROFESSIONAL VIEW fc AD1 RR-AME @

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June 30,1989 y

l NEMORANDUM FOR:. Frank J. Miraglia.. Associate Director for Inspection and Enforcement FROM:

Robert B. A. Licciardo, Reactor Engineer Plant Systems Branch Division of Engineering and Systems Technology F

SUBJECT:

DIFFERINGPROFESSIONALVIEW(DPV)CONCERNINGCONTAINMENT l-ISOLATION VALVES AT ZION On June 16, 1989, the writer did elaborate for the Standing Review Panet upon the principal regulatory positions sumarily presented in his OPY of May 11, 1989. Ne shall be please to clarify further on the specific issues identified in your memo to him of June 23, 1989, and will do so by July 17, 1989.

4b'M Robert B. A. Licciardo Registered Professional Engineer California Nuclear Engineering License No. NU 001056 Mechanical Engineering License No. M 015380 cc:

J. Sniezek C. Rossi-F. Congel H. Smith

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I wasHmotow,0. c. rossa July 14,1989 MEMORANDUM FOR:

Frank J. Miraglia, Associate Director for Inspection and Enforcement FROM:

Robert B. A. Licciardo, Reactor Engineer Plant Systems Branch Divisior, of Engineering and Systems Technology

SUBJECT:

DIFFERINGPROFESSIONALVIEW(DPV)CONCERNINGCCNTAINMENT ISOLATION VALVES AT ZION l

By inamo dated June 30, 1989, the writer proposed to submit requested clarifications of-his DPV by July 17, 1989. He would like to re-schedule this submittal to F-He is of course, prepared to agree to an extensionof.therede.;00. ired formal completion of the review of his DPV, by the same time per lod.

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f dowsr-Robert B. A. Licciardo Registered Professional Engineer California l-Mechanical Engineering License No. M 015380 Nuclear Engineering License No. NU 001056 cc:

J.. Sniezek C. Rossi F. Congel H. Smith II

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MEMORANDUM FOR: Frank J. Miraglia, Associate Director for Inspection and Enforcement f

FROM:

Robert B.'A. Licciardo, Reacto. Engineer.

Plant Systems Branch Division of Engineering and Systems Technology

SUBJECT:

DIFFERINGPROFESSIONALVIEW(DPV)CONCERNINGCONTAINMENT i

ISOLATION VALVES AT ZION By memo dated June 30, 1989, the writer proposed to submit requested j

L clarifications of his DPV by July 17, 1989. He would like to re-schedule I

l this submittal to July 20. He is of course, pre >ared to agree to an extension of the required formal completion of tie review of his DPV, by the same time period.

hd" Robert B. A. Licciardo l

4 Registered Professional Engineer California Nuclear Engineering License No. NU 001056 l

-Mechanical Engineering License No. M 015380 cc:

J. Snierek l

C. Rossi F. Congel l

H. Smith i

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' 9, July 21, 1989 MEMORAmptM FOR: Robert Licciardo, Reactor Engineer e

Plant systems tranch, DEST rRON:

Frank J. Miraglia, Associate Director i

for Inspection and Technical Assessment 1

$NECh DPV CONCERNING CONTAINMENT 150LAT10N VALVES AT ZION (TAC 73427). STATUS INFORMATION L

This is to acknowledge that on July 20,1989, I received the information requested by sy June 30,1989 memorandum. Based on the receipt of this information, the Standing Review Panel (E. Rossi, F. Congol and syself) can continue its review of your concerns. The Pane) expects to complete this review within 30 days, as provided for in Manual Chagter (MC) 4125 (see age 2, item 7 of the enclosure to MC 4125 entitled Procedures for the pression and Disposition of Differing Professional Views and Opinions

• which is attached to NRR Office Letter No. 300, Revision 1 dated March $4,1989).

Pursuant to MC 4125, the Director, hRR is expected to make a final disposition of the issue within seven days of the fanel's decision.

By copy of this memorandum NRR is notifying the E00 of the status of the Differing Professional View that you filed on'May 11,1989. The enclosed l

chronology documents the status to date.

i orastaansim oe W rrentJ.sams11a Frank J. Miraglia, Associate Director for Inspection and Technical Assessment

Enclosure:

Chronology DISTR!!UT ON cc w/encle centra r les V. Stallo, ECO ADMRF(2)

J. Blaha, A0, ED0 HSmith T. Murity VWilson l

J. Sniesek Jlarkins F. Gillespie FMiraglia '

E. Rossi F. Congol J. Partlow

4. Holahan A. Thadant C. Patel C. McCracken i'"Hi,,Mai 4Meef4*esebe.

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07/ O 89 I

,- 4 CHRONOLOGY Robert Licciardo DPV Concerning Containment 1 solation Valves at Zion l

Date Description 5/11/89 Memorandum from R. Licciardo to T. Murley, subject:

DifferinfiProfessionalView(DPV)Concerning(a)Issuanc?of 5ER to Z'on I and 2 allowing fuel-power operation with open 42" containment isolation valves; (b) methodology used for calculating related offsite doses, i

5/18/89 Memorandum from T. Murley to R. Licciardo acknowledging that on May 12, 1989, the Director's Office received his DPV dated May ll, 1989.

Note that this memorandum also requests a listing of persons Mr. Licciardo would like to be. considered for the third and alternate members for the Standing Review Panel.

i 5/25/89 Memorandum from R. Licciardo to T. Murley, nominating 5. Varga, F. Congel and G. Holahan to consider as third and alternate l

members of the Standing Review Panel.

5/26/89 Memorandum from T. Murley to F. Miraglia, C. E. Rossi, and F. Congel stating that they have designated the panel to review and recommend to the Director, NRR, the appropriate disposition of Mr. Licciardo's DPV.

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6/2/89 Memoranous from F. Miraglia to R. Licciardo acknowledging that the Standing Review Panel has detemined that adequate infonption hn been supplied to initiate a review of the DPV.

6/16/89 The Standihg Review Panel met with Mr. Licciardo to further discuss his views.

6/23/89 Memorandum from F. Miraglia to R. Licciardo acknowledges meeting held on June 16,1989, and the Panel's request for the following information: (1)requestthatMr.Licciardomore clearly state his concern regarding the time to fuel failure used in LOCA analyses; (2) request that Mr. Licciardo clarify the mechanisms for transporting fission products from the rimary to containment used in Mr. Licciardo's analyses; and l

l p(3) request that he provide his view as to the safety significance of proceeding with the proposed Zion amendment and the safety significance of his concern regarding LOCK analyses, t

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4 6/30/89 Memorandum from R. Licciardo to F. Miraglia stating that the information requested by memo from F. Miraglia dated 6/23/89 will be provided by July 17, 1989.

7/14/89 Memorandum from R. Licciardo to F. Miraglia advising that information provided by his memorandum dated June 30, 1989 will be submitted by July 20, 1989 7/20/89 Memorandum from R. Licciardo to F. Miraglia trensmitting information requested on June 23,1989.

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