ML17158A919
| ML17158A919 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 09/28/1995 |
| From: | Zwolinski J NRC (Affiliation Not Assigned) |
| To: | Blanch P AFFILIATION NOT ASSIGNED |
| References | |
| NUDOCS 9510040291 | |
| Download: ML17158A919 (10) | |
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UNITED STATES NUCLEAR REGULATORY COMMlSSION WASHINGTON, D.C. 20555-0001 September 28, 1995 Hr. Paul N. Blanch 135 Hyde Road West Hartford, Connecticut 06117
Dear Mr. Blanch:
I am writing in response to several letters from you regarding technical issues which you identify as having substantial safety significance.
Specifically, I am responding to your letters of January 26, February 28 and March 2, 1995.
In those letters, you raised concerns regarding pressure locking of gate valves and potential spent fuel pool cooling system design deficiencies.
The staff has reviewed the technical issues raised in your recent letters, including gate valve-pressure locking and spent fuel pool cooling issues, and concluded that recent and planned staff activities were adequate to address potential safety concerns.
In your January 26 and February 28, 1995, letters, you raised issues related to the consequences of a postulated spent fuel pool accident.
I recognize that the letters are related to the petition you filed, pursuant to 10 CFR 2.206, on April 13, 1994.
In your January 26, 1995 letter, you asked that the.
staff discuss how the October 24, 1994, draft safety evaluation on spent fuel pool cooling at Susquehanna addressed concerns identified in your 2.206 petition.
That document was not forwarded for the purpose of addressing specific issues raised in your petition.
Rather, it was provided for your-information as a matter of courtesy.
Staff consideration of your petition is continuing and the issues raised therein will be addressed in a Director's Decision.
However, your recent letters requested that the staff consider certain additional spent fuel pool cooling issues not included in your original petition.
A detailed discussion on your more recent spent fuel cooling issues, as well as the gate valve pressure locking issues raised in your Narch 2, 1995 letter, is presented below.
In thy January 26, 1995 letter, you asked that the staff (1) provide a copy of NRC procedures for conducting safety evaluations, and (2) perform an offsite radiological dose calculation for a certain postulated spent fuel pool accident during refueling.
In a letter dated February 28, 1995, you clarified these requested actions and also requested to know (1) if, given a postulated spent fuel pool accident, a licensee would remain within the offsite dose consequence guidelines cited in 10 CFR Part
- 100, (2) additional details on offsite dose consequence parameters, and (3) clarification on the land use impact and property damage assessment provided in NUREG-1353.
Before discussing the technical issues you raised, I want to respond to your request for NRC procedures For "conducting continuing safety evaluations or for tracking generic safety issues."
Under current practice, the staff has a
number of processes for reviewing and screening potentially safety significant 951004029i 950928 PDR ADQCK 05000387 P
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P.
Blanch issues that arise from a variety of sources.
The staff is attempting to combine and formalize existing processes into one centralized process.
The goal of the staff's effort is to provide a more centralized process for consistent identification, evaluation, prioritization and documentation of the considerations involved in the resolution of the more significant generic issues.
In general
- terms, the process would provide that information which may contain potential safety issues from any source would be reviewed by a single organization.
Items identified as potential safety issues would be reviewed by a multi-disciplinary panel.
The panel would then review the need and priority for resolution activity and make additional recommendations as appropriate.
Each issue assigned for resolution would be tracked until resolution is complete.
Implementation of planned activities would be executed and tracked by appropriate NRC organizations.
The revised process would include, for the more important issues, documentation of those considerations used to identify the resolution activity.
Resolution of any issue may include issuance of appropriate generic communications.
You also requested that the staff perform a calculation of the offsite radiological consequences for a certain spent fuel pool accident postulated to occur during refueling.
In raising this issue, you stated:
Hy overwhelming concern is an accident during refueling that causes a loss of water in the spent fuel pool.
The initiating event for this scenario may be a seismic event, station blackout or a fuel handling accident that causes a rupture of the spent fuel pool liner or associated systems...
The public has a right to know the consequences of this low probability event.
The staff has analyzed the radiological consequences of various spent fuel pool accidents and made the results publicly available in NUREG-1353, "Regulatory Analysis for the Resolution of Generic Issue 82,
'Beyond Design Basis Accidents in Spent Fuel Pools'," April 1989.
The staff provided a copy of this document to you during your drop-in visit on March 3, 1995.
The radiological consequence value of 8.0xl0 person-rem to a population within a 50 mile radius for spent fuel pool accidents, which you cited in your January 26 letter, is presented in NUREG-1353 and was indeed quoted in the staff's safety evaluation of issues related to spent fuel pool cooling at Susquehanna.
However, in NUREG-1353, the staff presents the detailed assumptions and scenarios for which the value you quoted is germane.
The spent fuel pool accident in NUREG-1353 to which the 8.0xl0 person-rem was ascribed involves a Zircalloy cladding fire after a complete loss of spent fuel pool coolant inventory initiated by a beyond design basis seismic event.
NUREG-1353 provides a detailed discussion of the radiological consequences, including property damage issues such as those raised by you in your February 28, 1995 letter.
NUREG-1353 discusses
- other, less severe potential causes of a complete inventory loss including seal leaks and liner leaks which you mentioned in your letter and loss of spent fuel pool cooling events.
- However,
P. Blanch unlike your recent letters, the NUREG is careful not to convey the impression that any seal leak, liner leak or interruption of fuel pool cooling necessarily leads to the estimated offsite consequences.
- Instead, the NUREG provides a well-developed discussion on the estimated frequency of fuel pool accidents that places the estimated consequences in appropriate context.
You should also note that, as far as the specific calculations you requested, the NUREG presents a table of worst case consequences based on shorter spent fuel decay time and a high surrounding population density for the beyond design basis seismic event.
Because of the extensive radiological consequence analysis conducted for the development of NUREG-1353, the staff does not plan any additional consequence analysis to address the issues raised in your January 26 and February 28, 1995 letters.
I urge you to review NUREG-1353 in detail to understand the scope and the context of the staff's analysis.
In your February 28, 1995 letter, you requested to know whether a plant experiencing a complete uncovering of the spent fuel would be in compliance with the guidelines in 10 CFR Part 100.
Let me assure
- you, 10 CFR Part 100 plays an important role in establishing the siting and design of nuclear power plant facilities.
With regard to the fission product release to be used in determining site suitability, Part 100 states:
The fission product release assumed for these calculations should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental
- events, that would result in potential hazards not exceeded by those from any accident considered credible.
Such accidents have gener ally been assumed to result in substantial meltdown of the core with subsequent release of appreciable quantities of fission products.
Compliance with Part 100 is confirmed at the time of issuance of the facility operging license.
A body of regulatory guidance was developed over time that defined the spectrum of accidents, either "hypothesized for purposes of site analysis or postulated from considerations of possible accidental events,"
which were to be analyzed as a basis for concluding that a plant met the dose guidelines of Part 100.
This guidance is contained in the NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports of Nuclear Power Plants,"
(SRP) which contains reference to additional guidance documents (e.g.,
Based on information provided by the applicant and using the review guidance in the SRP, the staff makes a finding that a facility is in compliance with Part 100 at the time of issuance of the facility operating license.
Part 100 does not contain provision for the re-review of facility design and operation for issues outside those that formed the basis for the initial finding of compliance.
However, the NRC staff does use the dose guidelines of Part 100 as acceptance criteria in evaluating issues that may affect design basis accident dose consequences at operating reactors.
The staff does not perform, nor require
P.
Blanch the licensee to perform, a revised design basis accident dose consequence analysis for each licensing action or potential safety issue that is identified.
In the case of typical licensing actions and potential safety
- issues, the licensee will perform an analysis, of the issues against the assumptions in the original siting analysis.
If the licensee concludes that the assumptions in the original siting analysis remain valid, the licensee does not perform, nor does the staff require the licensee to perform, a
complete reanalysis of the offsite dose consequences.
Certain issues
- however, due to their nature or complexity, warrant a reanalysis of the offsite dose consequences.
In these
- cases, the licensee will perform, or the staff will request that the licensee
- perform, an analysis of the offsite dose consequences of the licensing action or the particular safety issue.
In a letter dated March 2,
- 1995, you requested that the staff perform offsite radiological consequence analyses for postulated core and containment failure that may result from containment sump recirculation valve pressure locking phenomenon at the Millstone Nuclear Power Plant, Unit 2.
You requested that the staff address compliance with Part 100 as part of its response.
You further requested that staff analyses address the impact on operation of Millstone Unit 1 and Unit 3.
In addition, you requested an explicit statement regarding compliance with all NRC regulations, especially those related to motor-operated
- valves, at Millstone.
Finally, you asked whether or not the Millstone licensee had concluded that every motor-operated valve at Millstone was operable.
With regard to your specific questions about the Millstone 2 facility, a special NRC inspection was conducted at Millstone, Unit 2, between February 6, 1995 and March 15, 1995 to evaluate the reported inoperability of the containment sump recirculation valves.
The inspection report was forwarded to you by letter dated Hay 3, 1995.
The NRC held an enforcement conference with the Millstone 2 licensee on April 18, 1995.
By letter dated Hay 24,
- 1995, the NRC issued a Notice of Violation and Proposed Civil Penalty in the Amount of $50,000 to Northeast Nuclear Energy Company for failure to take appropriate corrective action for conditions adverse to quality and for failufe to properly review and accept purchased material that would have led to identification of the potential pressure locking issue considerably earlier than it was identified.
In issuing the Notice of Violation and Proposed Civil Penalty, the NRC acknowledged the safety significance associated with the potential loss of the safety injection and containment spray cooling functions.
While the NRC will continue to follow the resolution of this issue at Millstone 2, the staff considers that the licensee has taken necessary steps to assure that the containment sump recirculation valves will retain their function and that as a
result the safety injection and containment spray cooling functions will also be retained.
As a result, the staff is satisfied that the radiological consequences of design-basis accidents will remain within regulatory limits.
P.
Blanch On a generic basis, as you noted in your letter, the staff issued Information Notice (IN) 95-14, "Susceptibility of Containment Sump Recirculation Gate Valves to Pressure Locking," on February 28, 1995.
In addition, the
- NRC, through its Regional offices, polled pressurized water reactor (PWR) licensees to determine which PWRs were also potentially susceptible to pressure locking of containment sump recirculation valves.
For facilities identified as being susceptible, the staff implemented a temporary inspection instruction (TI 2515/129, "Pressure Locking of PWR Containment Sump Recirculation Gate Valves" ) to review short term actions by affected licensees.
Finally, the staff has issued Generic Letter (GL) 95-07, "Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves,"
on August 17, 1995.
The GL addresses the overall issue of pressure locking and will ensure that this phenomenon is appropriately addressed for all safety related power-operated gate valves.
The staff considers the time allowed in the generic letter to ensure that licensees have addressed pressure locking and thermal binding to be appropriate based on previous staff and industry communications regarding this issue and the low probability of an event that could create a demand on the subject valves.
Nevertheless, where a licensee determines that a valve is susceptible to pressure locking or thermal binding, the licensee must take appropriate corrective a'ction which may include evaluation of the operability of the valve and action in accordance with its Technical Specifications, regardless of the generic letter schedule.
With respect to motor-operated valve operability, the staff has a program in place to conduct a series of inspections at every nuclear power plant to ensure that licensees are implementing programs to meet the guidelines of Generic Letter 89-10 and its supplements.
Licensees are obligated to meet the commitments they made in response to Generic Letter 89-10 and the staff will review licensees programs for compliance with the licensing basis, including appropriate regulations and commitments.
Additional information on operability will be provided under separate cover.
0
P.
Blanch Septenber 28, 1995 Let me conclude my response to your letters by emphasizing that the staff mission is always focussed on ensuring the health and safety of the public, with regard to the operation of the nation's licensed nuclear power plants.
The staff does not intend to per'form the dose calculations you have requested
- because, in my estimation, the staff resources can be used far more meaningfully by pursuing the courses of action developed by the staff for these two issues.
In implementing the actions described
- above, the staff is addressing the issues in a manner commensurate with their safety significance.
Should you wish to discuss these matters further, please do not hesitate to contact me at (301) 415-1453.
Sincerely,
/s/
John A. Zwolinski, Deputy Director Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
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P. Blanch Let me conclude my response to your letters by emphasizing that the staff mission is always focussed on ensuring the health and safety of the public, with regard to the oper ation of the nation's licensed nuclear power plants.
The staff does not intend to perform the dose calculations you have requested
- because, in my estimation, the staff resources can be used far more meaningfully by pursuing the courses of action developed by the staff for these two issues.
In implementing the actions described
- above, the staff is addressing the issues in a manner commensurate with their safety significance.
Should you wish to discuss these matters further, please do not hesitate to contact me at (301) 415-1453.
John
. Zwolinski, Deputy Director Divisson of Reactor Projects - I/II Office of Nuclear Reactor Regulation
DISTRIBUTION:
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