ML20046C763

Provides Comments on Draft NUREG-1465, Accident Source Terms for Light Water Nuclear Power Plants.

ML20046C763
Person / Time
Site:	Arkansas Nuclear
Issue date:	10/28/1992
From:	Mcgaha J ENTERGY OPERATIONS, INC.
To:	Soffer L NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
Shared Package
ML20046C749	List: ML20046C759 ML20046C763 ML20046C783
References
RTR-NUREG-1465 NUDOCS 9308120070
Download: ML20046C763 (7)

Text

. - - - . . . . . - - - . .

~.

1.

'= = ENTERGY EE".E"""T" JaC=sm MG 39'J84 Ph5 -

Tc G01984 97 30

~~~~

John R. McGaha '

iwi October 28,1992 Dr. I.eonard Soffer Division of Safety issue Resolution Office of Nuclear Regulatory research U.S. Nuclear Regulatory Commission -

Washington, D.C. 20555 ATTENTION: Docketing and Service Branch

Subject:

Entergy Operations Comment on Draft NUREG-1465, " Accident Source Terms for Light-Water Nuclear Power Plants"

Reference:

55 Federal Register 33374, dated July 28,1992 CNRO-92/00098

Dear Dr. Soffer:

The referenced Federal Register requested comments on the subject draft NUREG.

Entergy Operations, Inc., the licensee for Arkansas Nuclear One, Grand Gulf Nuclear Station, and Waterford 3 Steam Electric Station wishes to offer the following.-

The current effort to arrive at a more realistic source term is a worthy undertaking in light of advances over the past decade, We believe the industry's perspective is best -

. represented by the Electric Power Research Institute (EPRI) Advance'd Light Water -

Reactor (ALWR) Utility Requirements Document (URD) Program and Utility Steering Committee.' As such, we endorse their comments on this NUREG. In addition, Entergy . -

Operations'provided input to and endorses the comments submitted by the Nuclear Management And Resources Council, Inc. (NUMARC) on this NUREG.

We agree with the NRC Staff that the future plants provide the impetus for implementation of the revised source term. However,' Staff suggestions'that current licensees may voluntarily consider application of the revised source term should be - -

strengthened to commit NRC resources to a timely review of submittals.' Further, it may be appropriate to consider generic approval of some applications for existing licensees as this issue develops further.

9308120070 930722 PDR NUREG 1465 C PDR ,.

y .,,.,e-. -

,w. +-

Entergy Operations Comments on Draft NUREG-1465 CNRO-92/00098 Page 2 of 2 October 2'7,1992 There are several aspects of the new source term that are worth noting for existing licensees. These points are noted in the attachment to this letter, in particular, the more realistic timing for release of the source term could be of significant benefit to existing licensees if used with the original source term. This approach would allow for more realistic containment and control room isolation times, diesel start times, and other things, without necessitating significant and potentially cost prohibitive re analysis, in summary, we welcorne the NRC's undertaking to develop a more mechanistically correct source term for future plants and for existin0 Pl ants, on a voluntary basis. We l encourage the NRC staff to take action on the specific technical issues raised by EPRI's ALWR Utility Steering Committee in addition to our attached comments.

If you have any questions, please contact John Wilson at (601) 984-9756. Thank you.

Sincerely, 2)%

6 JR. A %

JRM/jkw attachment cc: Mr. R. P. Barkhurst Mr. J. J. Fisicaro Mr. J. W. Yelverton Mr. R. F. Burski Mr. W. K. Hughey Corporate File [ ]

l Mr. W. T. Cottle Mr. L. W. Laughlin DCC (ANO)

Mr. J. G. Dewease Mr. M. J. Meisner Records Center (W-3)

Central File (GGNS)

Attachment to CNRO-92/00098 Entergy Operations Comments on Draft NUREG-1465 Page 1 of 5 October 27,1992 Entergy Operations Comments on Draft Report NUREG-1465," Accident Source Terms for Light-Water Nuclear Power Plants"

GENERAL COMMENT

S Significant benefit could be gained by existing licensees if the more realistic timing for releases can be applied with the original source terms. These benefits include among other things:

. Relaxed containment isolation valve closure times. This could facilitate the use of more appropriately sized motors, which in turn would reduce associated valve wear caused by the harsh closure requirements. I i

. Increased margin for containment and control room isolation.

. Improved diesel reliability due to relaxed diesel start and load shedding times. This l would allow for a more rationale deliberate loading of important emergency core cooling systems.

If existing licensees are forced to use the revised source term to take advantage of these benefits, there should be the flexibility to apply the new source term for consideration of specific engineering applications rather than requiring whole scale re analysis for an entire plant.

The Staff should commit to assigning a reasonable priority to review applications to credit realistic timing and revised source terms for current licensees. It may also be appropriate to consider generic approval of some applications for current licensees as these applications are developed.

SPECIFIC COMMENTS FOR DESIGN BASIS EVENTS

1. Adoption of NUREG-1465 source terms as a successor to Regulatory Guide 1.3 and 1.4 could impact offsite and control room dose analyses for the loss of coolant accident (LOCA)in FSAR Chapter 15. Different aspects of NUREG-1465 could have both a positive and a negative impact upon the calculated dose

Attachment to CNRO-92/00098 Entergy Operations Comments on Draft NUREG-1465 Page 2 of 5 October 27,1992  ;

consequences. The magnitude of the impact would also be highly dependent upon interpretation of the NUREG-1465 for an FSAR design basis event vice a beyond design basis severe accident.

If NUREG-1465 is to be a successor document to Regulatory Guides 1.3 and 1.4, additional explicit guidance should be provided concerning percentages of lodine plateout on containment surfaces, revised spray removal coefficients for lodine species and other radionuclides in containment vapor, and other items currently addressed in the Reg Guides and other current regulatory guidance and requirements. .

2. We believe the source term based on NUREG-1465 should correspond to the Gap Release term and a fraction of the Early in-Vessel Release term for FSAR-type design basis events, since vessel breach does not occur for the FSAR design basis LOCA.

3. The NRC should consider adding justification to NUREG-1465 which addresses whether its conclusions, particularly concerning chemical forms of lodine, apply to FSAR-type design basis events. The information presented in NUREG-1465 is heavily derived from Severe Accident Sequences, and differences may exist in radionuclide characteristics (e.g., distribution in terms of species, chemical forms) under the different conditions.

SPECIFIC COMMENTS for IPE/PRAs Section_.1J We agree that some of the most severe releases arise from some containment I bypass events, such as rupture of multiple steam generator tubes. However, IPE studies have shown those containment failures occur very infrequently. l Without that caveat, the effect of a containment bypass is overstated.

in addition, NUREG-1465 states that source term assessments based on WASH-1400 are prevalent in PRA studies. In fact, NUREG/CR-4551 type assessments are more common in PRAllPEs. i Section 1.2  !

NUREG-1465 presumes that the vessel breach hole leaves a gaseous pathway l

~ . . - .. - - - . -

Attachment to CNRO-92/00098 Entergy Operations Comments on Draft NUREG-1465 Page 3 of 5 October 27,1992  ;

i from the reactor cavity to the upper containment atmosphere. It is that gaseous pathway that provides for flushing re-volatized fission products out of the RCS late on the accident time line. NUREG-1465 makes no allowance for the fact that some PWRs will have a submerged vessel breach following a large break loss of coolant accident (LBLOCA). Furthermore, NUREG-1465 does not allow for the vessel breach being delayed by ex-vessel cooling. Thus, in the

" conservative case" of the LBLOCA, late in-vessel releases do not really occur in tight, flooded reactor cavity plarits.

Further in this section, NUREG-1465 explains what high pressure melt ejection (HPME) is. However, this explanation should account for the vessel breach occurring under water. HPME at plants where the vessel is submerged results in very little aerosol reaching the upper containment. HPME is more dangerous for severe accidents like inter system loss of coolant accident (ISLOCA) and steam generator tube rupture (SGTR) which leave the reactor cavity dry. The cavity geometry also affects the amount of aerosol reaching the containment atmosphere. Depending on the core geometry, the corium may need to make two right angle turns before it could disperse into the containment proper. The NUREG should be clear that licensee's have the flexibility to consider their. ,

specific cavity geometry for source term dispersion.

Section 1.2 ends by describing steam explosions caused by corium dropping  !

into water. The description mixes in-vessel steam explosion phenomena with ex-vessel phenomena. The authors need to write two separate paragraphs on steam explosions, one for in-vessel and one for ex-vessel.

Section 3.2 NUREG-1465 should explain why, for a six inch break, that the time until the first fuel rod fails exceeds 6.5 (B&W) and 10 (W) minutes. The authors should clarify that their sensitivity case did not run any longer than those two times. They  ;

should extrapolate and explain whatever trend they saw.

Paragraph 5 of NUREG-1465 in this section says:

During the early in-vessel release phase, the fuel as well as other structural materials in the core reach sufficiently high temperatures b 'at the reactor core geometry is no longer maintained and fuel and other i

Attachment to CNRO-92/00098  !

Entergy Operations Comments on Draft NUREG-1465  ;

Page 4 of 5 l October 27,1992 i l

maten'als melt and relocate to the bottom of the reactorpressure vessel. l Those remarks need to consider the freezing at the lower core support plate that occurred at TMI-2. It is incorrect to say that the core always reaches the lower plenum.

l Later in the same paragraph the authors say:

This release phase ends when the bottom head of the reactor pressure vessel fails. ,

The timing described by that sentence assumes that ex-vessel cooling has no effect on the probability of vessel breach. Intuitively, ex-vessel cooling must I have some effect on vessel breach timing. In the LBLOCA severe accident l assumed throughout NUREG-1465 some PWRs' lower heads are submerged and may not breach.

Section 3.4 1

NUREG-1465 claims that a significant fraction of the sequences examined, in terms of frequency, occurred at low pressure. However, we believe most PWR results are dominated by high and medium pressure severe accidents starting off as station black out (SBO), SGTR, and small break loss of coolant accident (SBLOCA).

Tables 3.9.3.10 3.11.& 3.12 l l

It is not clear where the values for these tables are taken from. Those tables $

need to refer to the CURCOR terms from NUREGICR-4551 like FCOR, FCCID, and FLATE. The "Early in-Vessel column on Table 3.10 should correspond to the FCOR median number in the Surry part of NUREGICR-4551, Appendix B, I page B-2. Table 3.10 does not agree with page B-2 values for FCOR. Because j the relationship between numbers on Table 3.10 and terms like FCOR are not stated, the validity of the Table 3.10 numbers cannot be determined.

I NUREG-1465 does not provide a reference for Tables 3.11 and 3.12. The numbers on Table 3.12 do not match the Surry numbers on page B-2 of Appendix B for instance. It is not clear where these values come from.

i

Attachment to CNRO-92/00098 Entergy Operations Comments on Draft NUREG-1465 Page 5 of 5 October 27,1992 1

Section 4.1 The NUREG states "a high estimate" of fission product release to containment is selected, based upon a complete core melt scenario. For PRAs, an appropriate source term would be a realistic best estimate term based on a frequency-weighted average for severe accidents. For FSAR design basis analyses, a source term based upon complete core melt may be excessively conservative.

The NRC should consider providing source terms for the two or three scenarios which should be addressed. If only one source term is to be developed, a technical justification of why it is appropriate for the other conditions should be supplied; this should include justification that any single source term is not excessively conservative for the other applications.

As it stands, NUREG-1465 constructs source term estimates based on LBLOCA.

However, most severe accidents in IPEs arise from SBLOCA, ISLOCA, SGTR, SBO, and loss of all feodwater, not LBLOCA. An average source term for PRAs needs to be based on the most likely severe accident, not the most limiting one.

Section 4.5 The NUREG should explain here why more ionic iodine forms in PWRs than +

BWRs.

i