ML17180A850
| ML17180A850 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Quad Cities, LaSalle  |
| Issue date: | 07/25/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17180A848 | List: |
| References | |
| REF-GTECI-A-47, REF-GTECI-SY, TASK-A-47, TASK-OR GL-89-19, NUDOCS 9407280103 | |
| Download: ML17180A850 (4) | |
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e UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Enclosure l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THE BWR OWNERS GROUP RESPONSE TO GENERIC LETTER 89-19
- 1.
DISCUSSION On September 20, 1989, the NRC staff (hereafter referred to as staff), issued Generic Letter (GL) 89-19 regarding reactor vessel overfill protection. For BWRs, GL 89-19 discusses modifications to prevent a potential core melt event that bypasses containment.
The probability of core melt is very low, but the potential consequences can be significant. As a result, GL 89-19 reco1t111ends that all BWR plant designs provide automatic reactor vessel overfill protection to mitigate main feedwater overfill events.
The GL states that the design for the overfill protection system should be sufficiently separate from the main feedwater (MFW} control system to ensure HFW pump trip on a high water level signal in conjunction with a loss of power, loss of ventilation, or fire in the control portion of the MFW control system.
One of the base documents supporting GL 89-19, is NUREG 1218, "Regulatory Analysis for the Resolution of USI A-47," dated July 1989.
Chapter 4 of NUREG 1218 discusses possible General Electric BWR plant design changes.
The report communicates the NRC's recognition that the safety benefits gained by prQviding additional reactor vessel water level redundancy and independence to existing BWR overfill protection systems is not significant.
The report also states that modifying existing systems to provide additional channels is not a viable alternative in consideration of the cost/benefit cost analysis.
However, of the three plants that do not have automatic overfill protection capability, Oyster Creek is the only plant where modifications are warranted.
Subsequently, NRC approved the licensee's proposed design of automatic overfill protection system as. recommended in GL 89-19 to be installed at next refueling outage.
The remaining two plants are Lacrosse and Big Rock Point which are early vintage with low-power ratings and are located in low-density population areas.
The risk reduction for these two plants was estimated to be insignificant and therefore, modifications are not warranted.
Lacrosse has been permanently shutdown.
The staff also notes that Shoreham is permanently shutdown and is, therefore, not subject to GL 89-19 proposed actions.
In response to GL 89-19 and NUREG 1218, the BWR Owners Group (BWROG} submitted a report entitled "BWROG Response to NRC GL 89-19, "Hardware Change Recommendations," dated April 2, 1990.
The BWROG response was reviewed by Idaho National Engineering Laboratory {INEL) under contract to the NRC.
The results of the INEL review are documented by "Technical Evaluation Report:
Review of the BWR Owners Group Response to Reactor Vessel Overfill Protection; (Generic Letter 89-19)," dated February 1991.
The remainder of this Safety Evaluation is the staff's findings and conclusions based on its review of NUREG 1218, the BWROG response, and the INEL Evaluation.
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- 2.0 FINDINGS AND CONCLUSIONS This safety evaluation report (SER} is applicable to Millstone, Unit 1, and the BWR plants identified in NUREG 1218, the BWROG report and the INEL Technical Evaluation Report.
The staff reviewed the INEL Technical Evaluation Report (TER}, the BWROG sub-mittal, NUREG 1218 and BWR plant specific submittals.
Based on this review,
- the staff has concluded that it is highly unlikely that a loss of power event or a fire would Ciuse an overfill event by affecting the feedwater control circuitry and defeating the overfill protection since the feedwater control is an energize to act:wite system (e.g. the isolation valve will close upon loss-of-power).
The staff will confirm in the reviews of all plants that it is unlikely that any single event could disable overfill protection and the feedwater isolation.
Based on a comparison of the methodologies and the numeric results obtained in these documents, the staff concurs with the con-clusions and bases identified in the INEL TER.
The staff also notes that while the INEL evaluation includes no conclusion on bypass capability for the l-out-of-1 and l-out-of-2 trip logic overfill protection systems, the existing bypass capability is considered to be acceptable by the staff and is unaffected by the resolution of LISI A-47.
The staff's findings are summarized as follows with the understanding that the TER provides the technical basis for the findings.
(1)
Upgrading BWRs with existing automatic reactor vessel overfill protection to the separation and independence criteria identified in GL 89-19, is not warranted based on the cost/safety-benefit analysis.
(2)
As stated in GL 89-19, the staff recommends the following items:
(a) that plant procedures and technical specifications, for all BWR plants with reactor vessel overfill protection, include provisions to periodically verify the operability of overfill protection and ensure that automat.iiG:. overfill protection is available to mitigate main feedwater overfeed events during reactor power operation, and (b) that all BWR plawts reassess and modify, if needed, their operating procedures and operator training to assure that operators can mitigate reactor vessel overfill events that may occur via the condensate booster pumps during reduced system pressure operation.
Principal Contributor: S. Rhow, HICB/DRCH 504-2826
. -Mr. D. L. Farrar
- OFC NAME DATE COPY during low pressure operation prior to startup from the Unit 2 September 1990 outage.
Please confirm to the staff, in writing, that these revisions have been made.
- 2.
ComEd committed to include provisions to periodically verify operability of reactor vessel water l~vel instrumentation and overfill protection logic where necessary.
Specifically, ComEd indicated that Dresden would implement a daily instrument check for overfill protection instrumentation and that Quad Cities would implement a daily instrument check and trip logic functional testing every refueling outage by August 1, 1990.
Please confirm to the staff, in writing, that these changes have been implemented and describe how the daily instrument check is equivalent to the typically monthly channel functional test.
- 3.
ComEd committed to propose new TSs which would include provisions to periodically verify reactor pressure vessel overfill protection operability, ensure overfill protection is operable during power operation, and specify Limiting Conditions for Operation prior to startup from the Dresden and Quad Cities outages which began September 1990 and October 1990, respectively. These TS changes have yet to be submitted to the staff.
Please provide your response to the above request for information and submit reactor vessel overfill protection TSs for Dresden and Quad Cities as agreed to with Mr. Pete Piet of your staff by October 1, 1994.
If you have any questions concerning this matter, please contact me at (301) 504-1333.
The information requested by the letter is within the scope of the overall burden estimated in GL 89-19, which was a maximum of 240 person hours per licensee response.
This request is covered by Office of Management and Budget Clearance Number 3150-0011, which expires September 30, 1994.
Enclosure:
As stated cc w/enclosure:
See next page LA: PDIII-2 CMOORE
- 7/20/94 YES/NO Sincerely, original signed by R. Capra for Anthony T. Gody, Jr., Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation Distribution:
Docket File NRC & Local PDRs J. Roe J. Zwolinski R. Capra C. Moore A. Gody, Jr.
OGC 0-15-8-18 K. Jabbour 0-14-H-25 J. Stang C. Patel PDII I-2 R/F B. Clayton RII I ACRS (10) P-315 J. Wermiel 0-8-H-3 S. Rhow 0-8-H-3 PM:PDIII-2 PM: POI II-2 PM:PDIII-2 BC:HICB D: POI II-2 AGODY:lm~ JSTANG
- CPATEL
- JWERMIEL
- RCAPRA ~0-'-
u -
7/20/94 7/20/94 7/21/94
/2.5/94 7 /2.5/94 7
YES/NO YES/NO YES/NO YES/NO (E$}NO Name:
LA74955.LTR
- Mr. D. L. Farrar *
- OFC NAME during low pressure operation prior to startup from the Unit 2 September 1990 outage.
Please confirm to the staff, in writing, that these revisions have been made.
- 2.
ComEd committed to include provisions to periodically verify operability of reactor vessel water level instrumentation and overfill protection logic where necessary. Specifically, ComEd indicated that Dresden would implement a daily instrument check for overfill protection instrumenta-tion and that Quad Cities would implement a daily instrument check and trip logic functional testing every refueling outage by August 1, 1990.
Please confirm to the staff, in writing, that these changes have been implemented and describe how the daily instrument check is equivalent to the typically monthly channel functional test.
- 3.
ComEd committed to propose new TSs which would include provisions to periodically verify reactor pressure vessel overfill protection operability, ensure overfill protection is operable during power operation, and specify Limiting Conditions for Operation prior to startup from the Dresden and Quad Cities outages which began September 1990 and October 1990, respectively. These TS changes have yet to be submitted to the staff.
Please provide your response to the above request for information and submit reactor vessel overfill protection TSs as agreed to with your staff for Dresden and Quad Cities by October 1, 1994.
If you have any questions concerning this matter, please contact me at (301) 504-1333.
The information requested by the letter is within the scope of the overall burden estimated in GL 89-19, which was a maximum of 240 person hours per licensee response.
This request is covered by Office of Management and Budget Clearance Number 3150-0011, which expires September 30, 1994.
Enclosure:
As stated cc w/enclosure:
See next*page Name:
LA74955.LTR Sincerely, Anthony T. Gody, Jr., Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation Distribution:
Docket File NRC & Local PDRs J. Roe J. Zwolinski R. Capra C. Moore A. Gady, Jr.
OGC 0-15-8-18 K. Jabbour 0-14-H-25 J. Stang C. Patel PDIII-2 R/F B. Clayton RII I ACRS (10) P-315 J. Wermiel 0-8-H-3 S. Rhow 0-8-H-3 BC:HICB ~ D:PDIII-2 RCAPRA 7
94 YES/NO