ML19309B744
| ML19309B744 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 03/05/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19309B741 | List: |
| References | |
| NUDOCS 8004070132 | |
| Download: ML19309B744 (7) | |
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION 3~
7,1 g WASHINGTON, D. C 20555
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v SAFETY ASSESSMENT 0FFICE OF NUCLEAR REACTOR REGULATION BIG ROCK POINT PLANT SAFETY ISSUES IDENTIFIED BY MS. BIER AND MS. JOHNS INTRODUCTION By letters of November 4,1979 and January 6,1980, Ms. JoAnn Bier and Ms. Shirley J. Johns identified eight safety issues related to continued
- operation of the Big Rock Point Plant.
Six of the seven issues identified in the November 4,1979 letter were repeated, with clarifications, in the letter of January 6,1980. An eighth issue was added in the January 6, 1980 letter.
The discussion below lists issue number 4 as stated in the November 4,1979 letter and the other issues as stated in the January 6, 1980 letter and our assessment of each issue.
For several of the items, we intend to provide additional discussion in a future supplement to this assessment.
DISCUSSION 1.
We demand that our school systems have workable, safe evacuation plans for our children and that all private citizens be informed of appropriate in-dividual evacuation actions.
Paragraph 50.34(b)(6)(v) of Part 50 of Title 10 of the Code of-Federal Regulations (10 CFR Part 50) requires' that each appli-cation for an operating license include plans for coping with emergencies.
Appendix E to 10 CFR Part 50 establishes minimum requirements -for those plans.
Part IV D of Appendix E requires
" procedures for notifying, and agreements reached with local, State, and Federal officials -and agencies for the early warning of the public and for-public evacuation or other protective measures should such warning, evacuation, or other protective measures become necessary or desirable, including identification of the principal officials, by title and agencies." The emergency _
plan currently approved by NRCy' requires notification of a number detter from D. L. Ziemann, AEC to R. B. Sewell, CPCo dated November 27, 1974.
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of government organizations including the local sheriffs, Michigan State Police and the Michigan State Department of
. Health in the event of a serious emergency. Local and State officials would be responsible for notification of local school systems and evacuation. if needed. Therefore, a plan which meets the minimum requirements of our regulations currently exists. -However, by letter of October 10, 1979, we required all power reactor licensees to submit upgraded emergency plans which are standardized in format and certain other aspects, are more detailed than previous plans, and provide for accomplishment of more stringent objectives than required by the existing Appendix E to 10 CFR Part 50. A draf t revision to the Big Rock Point Plan was. submitted October 29, 1979 and is under review by an Emergency Planning Review Team.
The estimated completion date for full im-piementation of a revised plan is June 1980. Based on:
(1) the existence of an Emergency Plan which conforms to our current regulations, and (2) the results so far of our review of the draft revision of the plan, we have not identified any deficiencies in emergency planning which are so significant as to require suspen-sion of the operating license.
2.
We demand accountability for all diffuser pieces, which if left within the reactor vessel could interfere with rod movement and cause flow blockage as in the partial meltdown at the Fermi I Plant in Detroit.
On April 20, 1979, during shutdown conditions, Big Rock Point personnel detected a vibration-type noise in the lower pres-sure vessel when the No. I recirculation pump was in service.
Because of an unrelated problem (a leak in a control rod drive housing) the reactor core was defueled and vessel in-ternals removed.
Subsequent inspection revealed that the No.
I recirculation inlet baffle plate was loose from its mounting brackets on the vessel wall. The recirculation inlet. baffle plates were not completely effective, and in 1963 a new skirt
. baffle was installed on the core support plate and completely encircled all the support tubes. The old baffle plates had been lef t in place and over the years the three hold-down bolts were worn through allowing the baffle plate to vibrate against the new skirt baffle.
The safety consequences of this failure were minimal since the plates are designed such that they can-not enter the core or constitute a flow blockage. During the 1979 outage, new diffuser plates were installed using larger hold down bolts and positive nut locking devices.
m.---c All the bolt ends and nuts from the old plates were accounted for and the reactor vessel was cleaned and inspected to assure that there were no other loose materials that could affect reactor operations.
3.
We demand that Consumers Power Company systematically and absolutely make the necessary design alterations in the reactor level vessel instrument system LERE09 & LSRE09 which initiates reactor scram, containment isolation and core spray actuations.
By Licensee Event Report 79-22 submitted to the NRC by letter of September 22, 1979, Consumers Power Company (CPCo) reported
-a potential deficiency in the water level instrumentation used for reactor scram and initiation of engineered safety features.
CPCo modified the instrumentation and submitted an evaluation of the acceptability of the modification to us by letters of October 23 and October 31, 1979. By Amendment No. 31. dated November 2,1979 (copy enclosed) we approved revised license Technical Specifications for the modified instruments. The Safety Evaluation accompanying that amendment addresses the acceptability of changes to the Technical Specifications and the acceptability of the modification made to the water level instruments..As explained in that document, we have concluded that the design alterations are acceptable and that no further actions are necessary.
4.
We demand that the biological shield be made sufficient to contain deadly gamma rays in the event of loss-of-coolant accident (LOCA),
area residents would be protected and plant personnel would be able to perform necessary functions to bring the plant _under control.
One of the Lessons Learned from the TMI-2 accident is that radiation fields resulting from contained radiation sources after an accident may make it difficult to effectively per-form accident recovery operations or may impair safety equip-ment. As a result, by letter of October 30, 1979 we asked nuclear power plant licensees to perform a design review of plant shielding by January 1,1980 and to implement needed changes by January 1,1981.
Consumers Power Company submit-ted the design review by letter. of December 27, 1979 and identified areas of the plant which would need additional shielding protection if NRC design criteria were to be met.
The NRC design criteria assumes a very severe accident with a very large radiation source term and assumes that stringent limits on radiation exposure to personnel would be met.
Because~of.the conservatism of the design requirements against-which the shielding acceptability must be evaluated, it is our judgenent _that a deferral of implementation until 1981 does not expose! plant personnel to a significant ' risk from a loss-of coolant accident or a greater risk to the public thn.
- previously evaluated if one should occur.
However, we are still evaluating the CPCo letter of' December 27,-1979 and would require more immediate actions if.further review indicates they are. warranted.
5.
We demand that repetitive malfunction of their containnent isolation valves CV/4096, CV/4097 be resolved.
- 6. - We demand that repetitive malfunctions of valves CV/4027, CV/4105, M0/7050 be resolved.
. Repetitive malfunctions have occured in several containment isolation valves. LValve CV/4097 isL a butterfly valve in-the supply line of the containment ventilation system.
The
- valve is a replacement valve installed in April 1974. Ex-cessive leakage through this valve was reported March 31, 1975, June 5,1975,- May 3,1976, July 2,1976, Feb-uary 1,
~1978, September 12, 1978 and February 1,1979. Our records indicate'that with the possible exception of one test,'the leak rate through the line during accident conditions would have been acceptably limited by another operable isolation valve (CV/4096) in the same.line.
In one instance (LER R0-12-76 dated July 2,1976) our readily available recordsLdo not indicate whether the leak rate through the line would have been acceptably low.
In each cas,c, CPCo took corrective action to bring the leakage back to 'within acceptable limits and ~after repetitive failures CPCo initiated a review with the vendor to bring about long term improvements.
We will continue to monitor the test results on CV/4097 to determine if additional corrective actions are needed.
Our records-do not indicate repetive failures of valve CV/4096.
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-5.
M0/7050 is a main steam isolation valve.
A failure of this valve to close was reported September 12, 1978. A previous failure ~was reported April 5,1973. CPCo ordered a new type of valve packing as a long term corrective action.
We will also continue to monitor the test results on this vaive 1
-to-assure that the corrective action taken is sufficient.
CV/4027 is an automatic. isolation valve in the reactor and v
fuel pit drain line. Leakage in excess of technical specification limits for this line was reported by ' licensee event - reports dated June 10, 1975, and September'27, 1978.
In each instance, Valve CV/4117 which is redundant to Valve CV/4027, was operable and would have prevented excessive leakage through the line. The valve seats of CV/4027 have been machined to reduce the leak rate and CPCo has committed to installation of new valve seats.
The petition of November 4,1979 included one valve, CV/4117 which was not mentioned in the petition of January 6,1980.
We have no record of ' failures on valve CV/4117. This valve was identified in several Licensee Event Reports noted above as the valve which provided redundancy to a valve with ex-cessive leakage.
CV/4105 is an air operated isolation valve on the demineralized water line inside containment. Our records do not indicate a repetitive failure of this valve.
7.
We demand evidence that the BRNPF could withstand the crash of a B-52 Bomber without disaster to surrounding environment.
The concern with overflight of the Big Rock facility by air-craft began in 1963, when the Air Force installed an air-craft tracking station at Bayshore, Michigan, which is located approximately five miles from the Big Rock Point Plant.
Following this installation the Air Force began training the tracking station personnel in the detection of approaching ai rcra f t.
Concurrently, the Air Force was training the Flight-crews in avoiding detection by the radar station.
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In~ the beginning'it appeared that the Air Force was using the Big Rock Point' Plant as a flight target, 'since'there were many close overflights.
Consumers Power Company management complained to the Atomic Energy Commission (AEC) regarding this matter, and an agreement was reached with the Air Force.
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at -that. time to discontinue the direct low level overflights.
Low level overflights in the near vicinity of the plant con-
'tinued.until 1970 when the Big Rock Point Plant insurer
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raised the insurance rates because of these training flights in the near vicinity of the plant.
At that time, the Consumers Power President, Janes H. Campbell, contacted Congressman Gerald Ford, requesting that these training flights in the near vicinity of the Big Rock Point Plant be discontinued.
'At about this same time, in January of 1971, a flight crashed into the Little Traverse Bay approximately two miles from shore and about two miles frcu the plant.
As a result of these events the Air Force est ablished a training corridor which misses the Big Rock Plan: by three miles.
Air force charts were also marked to shoi that overflights of the Big Rock Point Plant were "Off.imits" and all training flights were to be confined to.he corridor.
From that time until July.1979 no low level ov' rflights have been observed by plant personnel.
In July 19 9 a low level overflight was observed and a complaint was re istered by Consumers Power v
Cor.pany management.
The Air Force investigated and reported that the-overflight was made by a new crew who had not observed the "Off-Limits" notation on their charts.
To prevent a re-currence, the Air Force stated that restrictions on overflights would'also be added to the flight checklists.
We are continuing our review of the risk asociated with air-craft near Big Rock Point in the Systematic Evaluation Program.
We have receantly received from the U. S. air Forcae a revised analysis of the probability of an aircraft crash at Big Rock Point. The Air Force analysis indicated an extremely low pro-bability of a crash into the Big Rock Plant.
We are reviewing the Air Force analysis and the need for further measures to in-crease assurance of safety.
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, 8.' -We demand that minimum requirements as established by the Nuclear Regulatory Commission for the Fire Protection System be met.
By License Amendments No.17, dated March 6,1978 and No. 25, dated April 14, 1979, we issued license con-f ditions to assure that an acceptable level of fire protection is achieved at the Big Rock Point Plant.
These amendments added limiting conditions of operation and surveillance requirements to assure that existing fire protection-equipment is operable and to require that modifications be made on a time schedule specified
-in Amendment No. 25 to further _ enhance fire protection at the plant.
The staff ~ safety evaluations associated with these amendments summarize our considerations in imposing there limiting conditions.
It is our judge-ment that sufficient; measures have been taken to permit continued plant operation prior to full implementation of all identified improvements identified in License Amendment'No. 25.
The petition of November 4,1979 included one concern not repeated in the petition of January 6,1980. That item (concern number 4) is addressed below.
4.
We demand that all NRC requirements issued to Consumers Power Company regarding the Oyster Creek occurence [ sic] be implemented with no pro-posed changes, technical specifications or adninistrative control com-promises allowed.
Following a loss of feedwater event at Oyster Creek Nuclear Generating. Station on May 2,1979, we determined that Big Rock Point was susceptable to a similar problem and would require
- a change in the technical specifications appended to the license prior to startup from the 1979 outage.
Our evaluation indicated that two additional technical specifications were
- appropriate for Big Rock Point and these technical specifications were issued October 30, 1979 prior to plant startup. A copy of Amendment No. 30 which changed the technical specifications and a copy of the associated NRC Staff Evaluation are enclosed.
It.is our judgement that the changes made are appropriate for
-Big Rock Point and.do not constitute any compromise of safety.
E.-
Enclosures:
1.
License Amendment No. 30, dated October 30,-1979
.2.
License. Amendment No, 31, dated' November-2, 1979-
.l Date:. March 5, 1980
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