NUREG-1353, Responds to Technical Concerns Re Loss of SFP Water Inventory in 970103 & 13 Ltrs: Difference between revisions

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#REDIRECT [[NUREG-1353, Board Notification 89-003:forwards Listed Documents Re Spent Fuel Pool Accidents for Resolution of Generic Issue 82, Beyond DBA in Spent Fuel Pools, Including NUREG-1353]]
| number = ML20137Z544
| issue date = 04/15/1997
| title = Responds to Technical Concerns Re Loss of SFP Water Inventory in 970103 & 13 Ltrs
| author name = Zwolinski J
| author affiliation = NRC (Affiliation Not Assigned)
| addressee name = Blanch P
| addressee affiliation = AFFILIATION NOT ASSIGNED
| docket = 05000245
| license number =
| contact person =
| case reference number = REF-GTECI-082, REF-GTECI-NI, RTR-NUREG-1353, TASK-082, TASK-82, TASK-OR
| document report number = NUDOCS 9704250019
| package number = ML20137Z548
| document type = CORRESPONDENCE-LETTERS, OUTGOING CORRESPONDENCE
| page count = 7
}}
 
=Text=
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. Soe8640M k            **/                              April 15, 1997
!            Mr. Paul M. Blanch 135 Hyde Road West Hartford, CT 06117
 
==Dear Mr. Blanch:==
 
i I as writing to respond to the technical concerns described in your letters
:            dated January 3 and January 13, 1997, regarding a loss of spent fuel pool i              (SFP) water inventory. I have forwarded your concerns about statements by Mr. Wayne Lanning and other NRC officials to the NRC Inspector General for
,            disposition.
Regarding your request for additional analyses of postulated SFP events i              specific to the Millstone site, I believe that the staff's evaluations of SFP
:              safety have been focused, methodical, and well documented.            Consequently, I i
believe that a careful reading of the staff's reviews will highlight that the i              staff has thoroughly considered the issues you raised. The following i              paragraphs describe my basis for this statement.
I            As you observed in your January 3, 1997 letter, a purpose of the NUREG-1353 study (NUREG-1353, " Regulatory Analysis for the Resolution of Generic Issue 82 l              [GI-82), 'Beyond Design Basis Accidents in Spent Fuel Pools'") was to evaluate the level of safety associated .with high density storage of irradiated fuel and determine if the cost of any proposed enhancements to fuel storage facilities are commensurate with the increase in the level of safety provided l            by the modification. Many of the concerns you have expressed involve assumptions made during the performance of the regulatory analysis and how the
!              results of the analysis fit into the NRC's regulatory process.
I              Each SFP and related system at licensed reactors has an associated design i
basis that the NRC staff reviewed and approved during licensing proceedings.
The design basis consists of the information that identifies 'the function of 1
!              the structure or system and the ranges of values for important parameters for                j
;            which the structure or system is designed to perform its function. As used in                  -
!              the title to NUREG-1353, the term "beyond design basis accidents" refers to
,              classes or magnitudes of postulated events that were not used in the design
,              process to define the function of structures or systems nor to establish .the
;              ranges of values for important parameters. Exclusionofeventsinacertjin class of or a certain magnitude from the design process is not necessarily-                  '
,              indicative of their likelihood relative to other events, such as large break l              loss-of-coolant accidents (LOCAs), that were considered during plant design.
j              However, the Commission must ensure that the integrated capability that 2
evolves from the events used in the design process provides adequate
;              protection for public health and safety.
  ;            With regard to the statement about the reporting requirements of 10 CFR 50.72
:            in your letter dated January 3,1997, the reporting requirements mentioned
:              (i.e., 10 CFR 50.72(b)(ii)(A) and (B)) apply following the actua7 occurrence l            of an event or condition that places the plant in an unanalyzed condition or a            s l
9704250019 970415 PDR    ADOCK 05000245 NRC fnE CEnlTER                        0
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P. Blanch                                                          condition outside the design basis of the plant. The postulated occurrence of events or conditions involving functional capabilities greater or different than those considered in the design process or assumed parameter values outside the range of values used in the design process do not constitute reportable events or conditions pursuant to 10 CFR 50.72. However, Northeast Nuclear Energy, the licensee for the Millstone reactors, has submitted several reports pursuant to 10 CFR 50.72 and 50.73 about actual events or conditions related to the SFP that the licensee believed placed the plant in an unanalyzed condition or a condition outside the design basis of the plant (e.g., piping having the function of retaining pressure following exposure to the stress values imposed by a seismic event had not been analyzed for that capability, and irradiated fuel was transferred to the SFP with a decay time that was outside of the range of values used in the design evaluation of the SFP cooling system).
Nevertheless, analyses of postulated event sequences serve an important purpose in the regulatory process. Through these analyses, event sequences can be evaluated for their level of risk. Where the risk is significant, the NRC employs the backfit process of 10 CFR 50.109 to modify the design or operation of a nuclear power plant. The evaluation documented in NUREG-1353 determined that the level of risk from beyond design basis SFP accidents was low, but the evaluation also examined the potential of specific enhancements to improve the level of safety.
A number of SFP event analyses were documented in NUREG-1353 to support resolution of GI-82. Events evaluated included seismic events of a magnitude greater than that used in the design process for the SFP structure, pneumatic seal failures, inadvertent drain down due to pipe breaks or system misalignments, and extended loss of cooling and makeup events. The analyses documented in NUREG-1353 estimated the probable radiological consequences of these events and the frequency with which the event would progress to the state producing the estimated radiological consequence.
In response to your concerns about adequate consideration of full core off-loads, you should note that the staff did consider the size and timing of the fuel off-load for events where the frequency of occurrence was strongly influenced by these factors. For example, the evaluation explicitly considered the impact of a full core off-load occurring 5 days after reactor shutdown for extended lo:s of pool cooling and makeup sequences because the size and timing of the off-load have an effect on the probability of recovery.
Conversely, the size and timing of fuel off-load has negligible influence on the progression of events initiated by large-magnitude seismic and other low-probability external events. For example, SFP structural failures due to such events as missile strikes and aircraft accidents were assumed to progress to rapid cladding oxidation regardless of the :in and timing of the most recent off-load because the event was assumed to preclude coolant addition to the SFP. Because of the independence of the progression of externally initiated events from the size and timing of the m st recent off-load, the staff believes that the use of a more probable radioneclide inventory (e.g., a l
l
 
3
    ^ '
P. Blanch                              !
one-third core discharge at 90 days decay) rather than an extreme inventory (e.g., a full core discharge with a short decay time) provides a more accurate
!          assessment of risk.
With regard to the consequences of the various events analyzed, the staff calculated best-estimate and worst-case consequences for events progressing to rapid cladding oxidation. The best-estimate case assumed that one third of a        -
core ignited as a result of the particular event sequence. The best-estimate case calculated radioisotope ' inventory based on the one-third core being placed in the SFP 90 days prior to the event occurring. Finally, as you observed, the best-estimate case assumed a typical population density of 340 people per square mile.
In contrast, the worst-case consequence estimate assumed that (1) the event
:          progression involved the igniticn of the entire inventory of stored spent fuel (a full fuel pool consisting of multiple cores stored over the hfe of the 4
plant), (2) the radioisotope inventery included the inventory of he niost
;          recent one-third core discharge with 30 days decay plus all previou s discharges, and (3) the assumed population density was 860 people per square mile, which was based on the Zion facility - a high population der sity site.
The staff does not believe further analysis of the offsite consequences considering the radioisotope inventory of a recently off-loaded full core is warranted given the highly conservative nature of the existing worst-case estimate.
i It is also important to understand that the worst-case consequence estimates            i were carried over into the value impact analysis. The value impact portion of          ;
NUREG-1353 quantified cost and benefits for two alternatives that were judged          l likely to mitigate the sequences that resulted in spect "uel fires (i.e., the          l 1          use of low density racks or increased use of dry storage). Both the best-          l estimate and the worst-case consequence estimates were factored into these cost-benefit evaluations. The staff also quantified costs and benefits for improvements to SFP cooling systems. Those cooling system modifications                l considered were judged likely to have only limited impact on averted offsite dose because the modeling of loss of pool cooling events did not support the assumption of rapid cladding oxidation. Instead, the analyses of loss of pool i          cooling events assumed cladding failure, which does not result in significant offsite consequences even when the decay time is as short as a few days.
i The bases for your assertions that "the probabilities and consequences of this accident have increased significantly" and that "the risk is now 100 times i
that previously assumed" are not clear. However, it appears that they are attempts to reconcile the results of recent staff evaluations of SFP safety with the results and conclusions in NUREG-l'13.
Note that the postulated loss of spent fuel water inventory events quantified i      in NUREG-1353 were total loss of water events that resulted in ignition of the stored spent fuel. NUREG-1353 specifically stated that "... spent fuel pools are designed to preclude significant (a few inches) fuel mcovery due to the i
 
l P. B1anch                                        leakage..." as a result of seal failures, pipe breaks, and other leakage events considered during plant design. The analytical models used to support the analyses in NUREG-1353 indicated that near total fuel uncovery (i.e.,
water level more than 12 feet below the top of the fuel) would be necessary for the significant fuel damage and ignition that formed the basis of the severe consequences quantified in NUREG-1353.
As part of the recent Task Action Plan for Spent Fuel Storage Pool Safety (Action Plan), the staff examined the features incorporated in the design of every operating reactor SFP to prevent coolant loss. The staff determined that all SFP designs had been reviewed and accepted by the NRC staff and that the designs provided adequate protection against loss of coolant events.
However, certain plants had features inconsistent with current SFP design guidance that could be enhanced to further reduce the probability of significant coolant inventory loss. As described in the staff's report to the Commission on the Action Plan, which was provided to you on September 11, 1996, the staff is performing regulatory analyses for those plants in order to determine if a substhntial increase in the level of safety at a justifiable cost can be achieved by modifying the SFP design or operation at those facilities. No enhancements to SFP design features were identified at the Millstone units with regard to reducing the probability of a significant coolant inventory loss, but Millstone Unit I was identified for potential enhancement of SFP temperature instrumentation if it is justified on a cost-benefit basis.
The staff study, AE0D/S96-02, which you cited in your letter, further quantified the probability of SFP leakage events. AE00/S96-02 did state that events involving a loss of SFP inventory greater than one foot have occurred at a rate of about one per 100 reactor years. The report notes that, as a result of human interaction, all actual events were terminated with approximately 20 feet of water remaining over the top of the fuel. As you correctly point out, an estimate for the frequency of loss of pool level of greater than one foot is in no way equivalent to an estimate of the frequency of a postulated loss of pool level down to the top of the stored fuel much less a total loss of inventory. Because NUREG-1353 accounts for human intervention to mitigate loss of inventory events due to seal failures or inadvertent drainage, there is no reason to believe that the estimates for these events in NUREG-1353 are inadequate or are inconsistent with the findings of AE00/S96-02. Thus, the requantification of the events in NUREG-1353 based on the findings of AE00/S96-02 is not warranted for Millstone or any other site.
The staff believes that the SFP safety studies conducted over the past few years, including NUREG-1353, the Task Action Plan for Spent Fuel Storage Pool Safety, and AE0D/S96-02, have provided significant insight into the relative risks posed by the storage of irradiated fuel in storage pools at the nations's power reactor facilities. These studies have identified the issues and facilities where specific design and operational improvements may be justified by reviewing previous analyses, such as NUREG-1353, and new information in a methodical manner. At the same time, these studies provide
 
I P. Blanch                                ,
s l
part of the justification for the staff's statement in Partial Director's                        l Decision DD-96-23 that the safety significance of certain full are off-load practices at Millstone 1 was low. The staff's November 9, 1995, safety
:            evaluation supporting the license amendment related to the practice of full core off-loads was also used to develop DD-96-23.
I do not believe that performing a site-specific analysis of beyond design                      ,
basis SFP accidents, which would be necessary to answer many of the detailed                    l questions in your February 28, 1995, and January 3, 1997, letters, would yield                  i information that could be used by the staff to improve SFP safety. Rather, I believe that staff's efforts to complete the actions identified for resolution of the Action Plan represent the most effective means of ensuring accomplishment of the NRC's mission of protecting public health and safety with regard to SFPs. Additionally, I believe these actions represent a highly                    1 responsive and responsible approach to the concerns you have raised.
On an administrative note, in your January 13, 1997, letter, you stated that your request was being made under the provisions of the Freedom of Information Act (F0IA). The existing staff analyses for spent fuel storage pool safety are documented in NUREG-1353, the July 26, 1996, report to the Commission on the Action Plan and in AEOD/S96-02. NUREG-1353 was provided to you during a drop-in meeting with me in March 1995. The Action Plan report was forwarded                      l to you on September 11, 1996. From your letters, it appears that you already have access to AE00/S96-02. As noted above, the staff's November 9, 1995, safety evaluation supporting the licensee's full core off-load amendment was also used to develop DD-96-23. This document is publicly available. There are no other " safety analyses" that substantiate the staff's conclusions, therefore, there is nothing to provide to you under F0IA that you do not already have or have access to.
If you have any additional questions on this matter, please do not hesitate to contact me.
Sincerely,
                                                                /S/
John A. Zwolinski, Deputy Director Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
* Previous Concurrence n                [h                                                    ,
OFFICE    PdhhA      P bf-2/PM  PDI-2/D  NRR/SP0
* DSSA/D
* DRP NAME      Mdhib        Wh          JStolz    PMcKee        GHolahan    JZdok1hski DATE      Nfhf97      Qho/97      Y/h/97      3/27/97          4/01/97 h/ Mf97 0FFICIAL' RECORD COPY                FILENAME: G:\SHEA\ BLANCH.197
 
P. Blanch                            '
part of the justification for the staff's statement in Partial Director's Decision DD-96-23 that the safety significance of certain full core off-load practices at Hillstone I was low. The staff's November 9, 1995, safety evaluation supporting the license amendment related to the practice of full
,      core off-loads was also used to develop DD-96-23.
I do not believe that performing a site-specific analysis of beyond design basis SFP accidents, which would be necessary to answer many of the detailed questions in your February 28, 1995, and January 3, 1997, letters, would yield information that could be used by the staff to improve SFP safety. Rather, I believe that staff's efforts to complete the actions identified for resolution of the Action Plan represent the most effective means of ensuring accomplishment of the NRC's mission of protecting public health and safety with regard to SFPs. Additionally, I believe these actions represent a highly responsive and responsible approach to the concerns you have raised.
On an administrative note, in your January 13, 1997, letter, you stated that your request was being made under the provisions of the Freedom of Information Act (F0IA). The existing staff analyses for spent fuel storage pool safety are documented in NUREG-1353, the July 26, 1996, report to the Commission on the Action Plan and in AE00/S96-02. NUREG-1353 was provided to you during a drop-in meeting with me in March 1995. The Action Plan report was forwarded to you on September 11, 1996. From your letters, it appears that you already have access to AE0D/S96-02. As noted above, the staff's November 9. 1995, safety evaluation supporting the licensee's full core off-load amendment was also used to develop DD-96-23. This document is publicly available. There are no other " safety analyses" that substantiate the staff's conclusions, therefore, there is nothing to provide to you under F0IA that you do not already have or have access to.
If you have any additional questions on this matter, please do not hesitate to contact me.
g Sincerely, John    . Zwolinski, Deputy Director Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
 
                                                                    ?
4 v
letter to P. Blanch from J. Zwolinski, dated April 15, 1997.
  .DISTRIBUTIQN :
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