ML19257B904
| ML19257B904 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 01/16/1980 |
| From: | Jens W DETROIT EDISON CO. |
| To: | Benaroya V Office of Nuclear Reactor Regulation |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 EF2-47-547, NUDOCS 8001210203 | |
| Download: ML19257B904 (5) | |
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(313) 237 8003 (313) 649-7117 January 16, 1980 EF2 - 47,547 Mr. V. Benaroya, Chief Auxiliary Systems Branch Division of System Safety Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Cornission 7920 Norfolk Avenue Bethesda, Maryland 20014
Subject:
Enrico Fermi Atomic Power Plant. Unit 2 NRC Docket No. 50-341
References:
1.
Regulatory Guide 1.07, Draft 1, Revision 2, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Condi-tions During and Following an Accident" 2.
Letter W. H. Jens to V.
- Benaroya, EF2-47,214, December 12, 1979 3.
Meeting NRC/ Detroit Edison, December 12, 1979
Dear Mr. Benaroya:
Your office has requested Detroit Edison generate specific comments on draft Regulatory Guide 1.47, Revision 2, to aid in conpressing the schedule for issuance and implementation.
Edison responded to an earlier NRC request by submitting a written statement (Refer-ence 2) defining the very large financial and schedule impacts that would result from a requirement for full compliance with the draft guide as it is presently written. This impact sur: mary was trans-mitted to your office in a timely fashion and was intended for your use in assessing the impact of the draft guide on a plant which received a Construction Permit before Regulatory Guide 1.97 was in existence.
AdM-(ha oe r
'T M fjoof 58'/0 l774 421 80012102 03 RE66D
Mr. V. Benaroya, Chief January 16, 1980 EF2 - 47,547 Page 2 Edison licensing representatives attended the December 12, 1070 meeting on the draft guide convened by the NRC staff to discuss comments received to date.
During the meeting the Fermi 2 state-ments(Reference 2) were referred to several times.
The context of the NRC staff renarks indicated that Detroit Fdison had not learned anything from the events at TMI-2.
Nothing could be more to the contrary; Edison has analyzed the incident that occurred at TMI-2 in considerable detail and has imposed similar scenarios, including fuel failure, to the Fermi 2 design as part of the review process.
A number of these specific areas are addressed in NUREG-0578. and Edison has committed to implementation of the required changes.
Edison has designed the present post-accident monitoring system to meet PSAR commitments to the AEC and has upgraded the design during FSAR review. A portion of the instrumentation presently installed has been tested to verify operability well beyond the original purchase design basis.
Based on Edison's interpretation of the draft guide as presently writtan. Edison offers for the staff's consideration the following general comments on the guide:
Inclusion of Type D and E variable classifications greatly e
weakens the thrust of the guide since the additional information added is not essential for mitigating the consequences of the accident and could degrade the operator's ability to respond.
In addition, the criteria for selection of the particular variables is absent, thereby requiring categorical conformance to a general list of variables.
e Variables are classified as Type B or C sithout specific justification. Defining the function / protection required and allowing the designer to classify the specific variables as Type B or C would greatly enhance the implementation of the intent of the guide.
e Requirement that Type C variables meet single failure criteria is not justified for a variable which by definition indicates only a potential for breaching fission product barriers.
e Imposition of the most recent equipment qualification standards on backfit plants is not reasonable.
e Categorical requirements for the use of noninterruptible IE power for variable Types B and C is not consistent with the function being performed. Design criteria for these two classes of instrumentation should only require that the loss-of-offsite power be considered in the design with provisions made to obtain acceptable performance.
177'4 322.
Mr. V. Benaroya, Chie f January 16, 1980 EF2 - 47,547 Page 3 e The duration of time that a variable is required is not adequately addressed by this draft. The imposition of a requirement for qualification lifetimes twice that of present standards is excessively conservative and seems to be based only on the TMI experience and not the particular plant design. Qualification duration should be addressed on a plant unique basis.
e Requirements for instruments with excessively wide ranges to encompass all ultimately possible regimes of operation will create a functional degradation of the plant information for normal operation and post-accident conitoring.
The best instru-ment system would be one that is used for normal plant operation and also serves for accident mitigation.
The specific ranges in the guide should be relaxed to allow the use of the dual purpose instrumentation.
Although a detailed review of all of the BWR post-accident variables listed in Table 3 of the draft guide has not been completed specific comments generated during Edison's preliminary review have been summarized below.
Comments are sequentially ordered and identify both the variable and the specific comment judged to be appropriate:
o Core Exit Temnerature This variable is not of significant value in a BWR since it is not a direct indication of adequate core cooling Following a loss of all level instrumentation, core damage could be more readily determined and assessed by use of main steam line radia-tion monitors, primary containment radiation monitors. and primary coolant sampling and analysis, e Neutron Flux Typical BWR source range monitors will meet the functional criteria for this variable. but are not qualified. The use of external core detectors would adequately verify core shutdown and should be considered as an acceptable alternative.
o Reactor Pressure Use of a 2000 psi upper requirement on range exceeds the value required for monitoring even the ATWS event.
This instrument should be required to have only a range of 110 percent of design vessel pressure.
e Coolant Level in the Reactor Range requirement is well beyond that possible for current RWR designs, the existing system range is adequate. Meeting the range would require adding taps to the vessel.
1774 4123"
Mr. V. Benaroya, Chief January 16, 1980 EF2 - 47,547 Page 4 e Main Steamline Flow This variable should be deleted from the list.
Steam flow is of no specific value since the system automatically isolates on numerous accident signals. Typically, at least two auto-matica11y operated valves are provided in each line for isola-tion purposes; a steam flow measurement will not provide any significant additional information.
e Main Steamline Isolation Leakare Control System Pressure This variable should be classified as Type C only and indication need not be redundant since the system is functionally redundant.
e Primary Sa fety Relief Valve Position This variable requirement should be modified to conform with NUREG-0578.
The requirements for single-failure design and strip chart recording should be removed.
e Radiation Level in Coolant This instrument is not required, the main steam line radiation monitor provides adequate indication of fuel damage prior to isolation. Primary containment radiation monitoring and coolant sampling are adequate following the isolation.
e Primary Containment Pressure The range requirement is not compatible with the accuracy requirements for the various functional uses, a range of 110 percent of design will provide a functional instrument for the entire spectrum of uses.
e Containment Isolation Valve Position This variable should be classified as type D only.
Redundant indicators on each valve are not required since the isolation valves are redundant.
Strip chart recording of valve position is not reasonable; a status indication is adequate.
Noninterrupt-ible power should not be required for these indicators, automati-cally restored 1F power should be acceptable.
1774 324
Mr. V. Benaroya, Chief January 16, 1980 EF 2 - 47,547 Page 5 o Suppression Pool Water Level The proposed range requirements are not consistent with the requirement defined in NUREG-0578. This range does not apply specifically to the Mark I containment designs.
e Drywell Pressure This variable is the same as containment pressure for Mark I BWR designs and should be deleted, o Drywell Sumps This variable should not be classified as a Type B since the sumps are isolated during and following an accident.
e High Range Containment Area Radiation This variable should be classified as Type C since it provides only an indirect measurcment of fuel integrity. The guide should be -itten to allow the use of monitors external to the drywell wall; or, as an alternative the use of a sampling technique.
e Temperature of Space in the Vicinity of Fouipment recuired for Safety This variable should be classified as Type D since it is monitor-ing the performance of a safety (cooling) system and not the information required to monitor the accomplishment of a critical safety function.
e Environs Radioactivity This measurement is not required for a BWR since the secondary containment itself is a leakage barrier enclosing the primary containment in contrast to the PWR design.
Edison has generated this set of comments based on knowledge of the design basis of the Fermi 2 post-accident monitoring instrumen-tation. NRC acceptance and incorporation of these comments into the final revision of Regulatory Guide 1.97, Revision 2, would, in our view, improve the safety design basis of the proposed post-accident monitoring instrumentation.
Sincerely, 17 4
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e Wayne H. Jens Vice President - Fuclear Operations WHJ/EFM:jl