ML20054C393
| ML20054C393 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/07/1982 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | GENERAL PUBLIC UTILITIES CORP. |
| Shared Package | |
| ML20054C389 | List: |
| References | |
| ISSUANCES-OLA, NUDOCS 8204200502 | |
| Download: ML20054C393 (12) | |
Text
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e UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of METROPOLITAN EDISON COMPANY, et. al.
Docket No. 50-320 OLA (Three Mile Island Nuclear Station, )
Unit 2)
)
AMENDMENT OF ORDER 1.
GPU Nuclear Corporation, Metropolitan Edison Company Jersey Central Power
~
and Light Company and Pennsylvania Electric Company (collectively, the licensee) are the holders of Facility Operating License No. DPR-73, which had authorized operation of the Three Mile Island Nuclear Station, Unit 2 (TMI-2) at power a
levels up to 2772 megawatts thennal.
The facility, which is located in Londonderry Township, Dauphin County, Pennsylvania, is a pressurized water reactor previously used for the commercial generation of electricity.
By Order for Modification of License, dated July 20, 1979, the licensee's authority to operate the facility was suspended and the 1icensee's autnority was limited to maintenance of the facility in the shutdown cooling mode then in effect (44 F.R. 45271).
By further Order of the Director, Office of Nuclear Reactor Regulation, dated February 11, 1980, a new set of formal license requirements was incosed to reflect the post-accident condition of the 8204200502 820409 DR ADOCK 05000320 PDR
, a facility and to assure the continued maintenance of the current safe, stable,
- long-term cooling condition of the facility (45 F.R.11282). Although these requirements were imposed on the licensee by the Director's Order of February 11, 1980, the TMI-2 license has not been formally amended.
The requirements are reflected in the proposed Recovery Mode Technical Specifi-cations presently pending before the Atomic Safety and Licensing Board.
Hereaf ter in this Amendment of Order, the requirements in question are l
identified by the applicable proposed Technical Specification.
II.
By letters dated February 23, 1981, March 18, 1981, and October 6,1981, the licensee proposed changes to the Recovery liode Technical Specifications for Three Mile Island Unit 2 (TMI-2) to reflect current plant conditions.
Several categories of changes were proposed.
The proposed changes are summarized as follows:
A)
Deletion of Measurement Instrumentation and/or Surveillance Requirements for Equipment No Longer Required to be Operable.
1)
The NRC staff finds that the present instrumentation used to monitor the shutdown condition of the reactor such as pressure and temperature transmitters, reactor coolant system baron sampling, and two operable source range neutron instruments is adequate; therefore, the deletion of the operability requirement for all differential pressure transmitters for the reactor coolant system as stated in Table 3.3-9 is acceptable as proposed.
2)
The proposed Recovery Mode Technical Specifications eliminated require-ments with respect to the Reactor Building Emergency Spray and Core Flooding Systems, but Table 3.3-10 of the Proposed Technical Specifi-cations still contains operability requirements for the Core Flood Tank Level and Reactor Building Spray Pump Flow Instrumentation in addition
3 to associated surveillance requirements (Table 4.3.10).
The staff does not find the subject instrumentation essential to maintaining the reactor coolant system in its present safe condition and therefore finds the proposed change acceptable.
3)
Proposed Technical Specification 3.7.10.2 (d) and (e) requires that the Deluge System located at the Hydrogen Purge Exhaust Filter and the Reactor Building Purge Exhaust Filter be operable with automatic initiation of water spray to minimize the fire hazards associated with the charcoal filter train.
The charcoal has since been removed, there-fore the proposed deletion of the requirement to maintain the deluge systems in an automatic initiation status is approved.
The deluge system would be isolated, thereby placing it in a manual made and therefore preventing spurious activation signals.
The fire detectors in the area are not affected and will continue to provide audible alarms.
4)
The licensee has requested deletion of the requirements of proposed Technical Specification 3.3.3.8 which requires the fire detection instrumentation for the Balance of Plant Diesel to be operable as l
imposed in the February 11, 1980, Order.
By letter dated April 28, 1980, the licensee proposed to delete these operability requirements l
for the 80P diesel generators and the 13.2Kv transmission line, re-placing their capabilities with an existing 230Kv grid system.
This modification was approved in the August 11, 1980, Modification of Order.
Since the diesels are located outside of the restricted area fence, are l
not located near any vital equipment, and will no longer be required to l
perform a safety function, the NRC staff concurs with the licensee's proposal.
I
l
.'o
., 5)
Proposed Technical Specification 3.6.1.3 (b) presently requires the containment air locks to be operable with an overall leakage rate of i
less than or equal to 0.05La at Pa, 56.2 psig.
The licensee, however, i
has previously been granted an exemption from certain requirements of 10 CFR Part 50, Appendix J upon which Proposed Technical Specification 3.6.1.3 (b) is based.
Therefore the licensee has requested and we have approved the deletion of the subject section.
However, Sectiun 3.6.1.3 (a) has not been deleted.
The footnote indicating that surveillance requirement 4.6.1.3 (a) represents an exemption to Appendix J has also been deleted since Appendix J was recently modified (45 FR 62789) and now requires the same leak rate testing as is presently stated in Surveillance Requirement 4.6.1.3(a).
B) Deletion of Systems or Structures No Longer Required to be Operable or Maintain Their Original Design Criteria.
1)
The Proposed Recovery Mode Technical Specifications retained the Hydrogen Purge Cleanup System (proposed Technical Specification 3.6.4.3) to ensure the operability of the system in the event that purging of the containment building was approved by the NRC.
On June 12, 1980, a Commission Memorandum and Order and a Temporary Modification of License were issued which authorized Metropolitan Edison Company to conduct a controlled purge of the containment building using that system with established off-site dose limits.
The purgo began on June 28,1980 and was completed on July 11, 1980.
Subsequent permission to perform i
periodic purges using another system, the Reactor Building Purge System, was granted by NRC letter NRC/TMI-80-119 from J.T. Collins to R.C. Arnold, dated July 31, 1980.
The licensee has therefore requested that the
5 Hydrogen Purge Cleanup System, Section 3.6.4.3, be deleted from the proposed Technical Specifications and Surveillance Requirement 4.6.4.3, since its purpose was fulfilled upon the completion of tne June 29, 1980, purge and the use of the system will not be required to maintain the containment building in a safe condition for the remainder of the Recovery Mode.
Per the discussion in item (2) of the enclosed safety evaluation, the NRC staff finds that the deletion of Sections 3.6.4.3 and 4.6.4.3 acceptable.
2)
Proposed Technical Specification 3.7.10.1 requires that at least 3 of 4 high pressure water pumps be operable to ensure that adequate fire suppression capability is available to confine and extinguish fires in any portion of the plant where safety related equipment is located.
The licensee has proposed to modify this requirement so that only 2 of 4 high pressure water pumps would be required operable.
The Fire Hazards Analysis for TMI determined that only 2 of 4 high pressure pumps were needed to provide a required combined capacity greater than 3575 gpm.
The Bases for the Proposed Technical Specifications also reflect the same conclusion.
Therefore, the staff has concluded that enough conser vatism is present in the proposed Technical Specifications Section 3.7.10.1 to allow a change in the requirements to 2' of 4 high pressure water pumps being maintained in an operable condition without impairing the Fire Suppression System.
3)
The licensee proposed to add occupational exposure considerations to the requirement stated in Section 5.2.2 of the Proposed Technical Specifications to maintain the 60 psig maximum internal pressure and 286 F maximum internal temperature design.
This would limit 0
maintenance on the inside of the containment if radiological conditions posed a hazard to personnel.
The staff approves this additi.1
'., but also sought to more correctly state a reasonable containment design pressure that should be' maintained.
The licensee performed an analysis This evaluating the peak containment pressure under accident conditions.
analysis which was independen-ly verified by the NRC staff, concludes that the maximum potential containment building pressure resulting from a LOCA is approximately 2 psig.. Subsequent to the licensee's initial analysis, another evaluation was submitted by the licensee, (LL2-81-091, Barton to Snyder,12/4/81), which concluded that it is conceivable that a fire inside of the containment building could increase the internal pressure to greater than 2 psig, causing a failure of the most limiting penetration (R-626).
However, the fire analyzed by the licensee illus-trated that even with the failure of penetration R-626, offsite releases would be less than 400 pCi with an average concentration at the station vent calculated to be 1.2 X 10-9 uCi/ml.The licensee's calculations and the staff's independent evaluation concludes that the releases would be well below the limits of 10 CFR 20 and within the scope of, impacts ass in the " Final Programmatic Environmental Impact Statement Related to t Decontamination and Disposal of Radioactive Wastes Resulting from the Also to minimize the Accident at TMI-2", dated March 1981.
March 28, 1979, potential for a fire induced overpressure, the licensee has instit hazards controls for the containment and is currently re-writing appli In addition, piping and electrical penetrations other than procedures.
R-626 will remain as is until R-626 are being redesigned to 5 psig.
accessibility and personnel exposure permit its possible upgrading to th 5 psig value.
The staff's independent evaluation agrees with the licensee's conclusio and therefore Section 5.2.2 has been modified to state a con
design pressure of 2 psig.
The design temperature will remain at 286 0F.
C) Addition or Modification of System or Structure Requirements.
1)
The licensee had originally requested that the requirements of Section 3.6.1.4 of the Proposed Technical Specifications which limits maximum primary containment pressure to <0 psig be deleted.
The basis for this request was that since the reactor building purge has removed most airborne contamination t~ rom inside containment, to the extent that normal leakage from the building will not affect the health and safety of the public,the requirement could be relaxed. However, based on discussions with the NRC staff on additional inside containment activities that could increase the airborne radioactivity concentration.
the licensee withdrew the requested modification to increase the maximum operating containment pressure to + 1 psig per LL2-81-091, Barton to Snyder, dated December 4,1981.
2)
The proposed Technical Specifications do not require that the control room air inlet radiation monitor remain operable for all credible accident conditions.
Presently, a failure of the control room air inlet radiation monitor could prevent the automatic shifting of the control __.
room emergency air cleanup system to the recirculation mode in the event of an accident.
The licensee has proposed to add the inlet monitor to proposed Technical Specification 3.7.7.1 and require it to be operable.
The staff agrees that this chance enhances the safety of control room f
personnel and concurs with the modification.
1 3)
ANSI N18.1-1971, Section 4.4 is the currently referenced technical personnel qualification criteria in proposed Technical Specification l
6.5.1.2 (c).
Specified in Section 4.4 are reactor engineering and physics, instrumentation and control, radiochemistry, and the radiation i
9 '
protection disciplines. The licensee has proposed to expand the academic requirements of the Plant Operations Review Committee (PORC) and the Generation Review Comnittee to a broader area as discussed in ANSI /ANS-3.1-1978, Section 4.7.2 The proposed change to the qualifi-cations of the PORC and GRC members substitutes a basic requirement for a bachelor's degree plus three years of professional level experience in lieu of the existing requirement for only five years experience or, at most, a bachelor's degree plus one year of experience.
The staff is in agreement that by using the proposed criteria the qualifications.for PORC and GRC members will be higner than those presently required in the existing Proposed Technical Specifications and, accordingly we conclude that the proposed change is acceptable.
D)
The Clarification of Ambiguity in Proposed Requirements.
- 1) The " action" statement of Proposed Technical Specification Section 3.9.12 (a) states that "with the fuel handling building / auxiliary building air cleanup system inoperable.....".
However, it is not clear in this statement or statement 3.9.12 (b) that the fuel handling building and the auxiliary building air cleanup systems are independent of each other and that one of the two systems being inoperable should not affect any radioactive movements in the building solely associated with the other operating system.
The staff concurs with a modification to 3.9.12 (a) and (b) as requested by the licensee to eliminate any ambiguity 2)
Section 3.6.1.1 of the Proposed Technical Specifications presently contains an incorrectly used "and" which has the unintended effect of requiring that containment isolation be maintained when one containment isolation valve per penetration is open or inoperable by (a) at least one deactivated automatic valve secured in the isolation condition '1d
_g_
(b) at least one closed manual valve or blind flange.
It was intended that method (a) og (b) would suffice when a penetration isolation valve was inoperable.
Therefore, the licensee's request to so modify the specification is acceptable.
3)
Paragraph 4.1.1.1.j. of the Surveillance Requirements presently indicates that water filled tanks, the surge tank and the degassed water supply tank shall be sampled to verify that they contain borated water.
The proposed change by the licensee specifies wnat tanks are to be sampled by using more accurate system terminology and is, therefore, acceptable.
Also, Section 4.1.1.1.j 2 specifies sampling to confirm a dissolved gas concentration of less than 15 scc /kg of water.
An (*) and footnote has also been added as requested in order to more specifically state where the most representative sample should be taken for water being added to the reactor coolant system via the Standby Pressure Control System.
The staff's safety assessment of this matter is set forth in the concurrently issued Safety Evaluation.
This evaluation concluded, in material part, that 1
the amendment of order does not involve a significant hazards consideration and that there is reasonable assurance that the health and safety of the l
public will not be endangered by operation in the modified manner. Prior 1
public notice of the amendment of order was therefore not required and the amendment of order is effective upon issuance.
l It was further determined that the amendment of order does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
In lignt of :nis determination, it was concluded that tne instant action is insignificant from l
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the standpoint of environmental impact and, pursuant to 10 CFR g51.5 (d) (4),
that an environmental impact statement or environmental impact appraisal need not be prepared herewith.
III.
Accordingly, pursuant to the Atomic Energy Act of 1954, as amenced, the Director's Order of February 11, 1980, is hereby revised to incorporate the deletions, additions, and modifications set forth in Attachment A hereto.
For further details with respect to this action, see (1) Letter to 8. Snyder,.
USNRC, from G. K. Hovey, Met-Ed/GPU, Technical Specification Change Request No. 26, dated February 23, 1981, (LL2-81-0049); (2)
Letter to B. Snyder, USNRC, from G. K. Hovey, Met-Ed/GPU, Technical Specification Change Request No. 26, Addendum A, dated March 18,1981,(LL2-81-0055);(3)
Letter to B. Snyder, USNRC, from G. K. Hovey, Met-Ed/GPU, Technical Specification Change Request No. 26 Addendum B, dated October 6, 1981 (LL2-81-0229); (4) Modification of Order dated August 11, 1980;(5)
Letter to R. C. Arnold, Met-Ed, from J. T. Collins, USNRC, granting permission to perform periodic purges using the Reactor Building Purge System, dated July 31, 1980; (6)
Memorandum and Order dated June 12,1980;(7)
Order l
for Temporary Modification of License, dated June 12, 1980; and (8) the Director's Order of February 11, 1980.
All of the above documents are available for inspection at the Commission's l
,, Public Document Room,1717 H Street, N.W., Washington, D.C., and at the Comission's Local Public Document Room at the State Library of Pennsylvania, Government Publications Section, Education Building, Commonwealth and Walnut Streets, Harrisburg, Pennsylvania 17126 FOR THE NUCLEAR REGULATORY COMMISSION kWk $
Harold R. Denton, Director Office of Nuclear Reactor Regulation Effective Date:
April 7,1982 Dated at Bethesda, Maryland l
l
References:
Letter to Bernard J. Snyder, NRC, from G. K. Hovey, Metropolitan Edison 1.
Company, " Technical Specification Change Report No. 26", LL2-81-0049, February 23, 1981 2.
Letter to Bernard J. Snyder, NRC, from G. K. Hovey, Metropolitan Edison Company, " Technical Specification Change Request No. 26, Addendum A",
LL2-81-0055, March 18, 1979.
3.
Letter to Lake Barrett, NRC, from G. K. Hovey, Metropolitan Edison Company,
" Request for an Exemption from the Testing Requirements of 10 CFR 50, Appendix J", LL2-81-0094, May 11,1981.
4.
Letter to G. K. Hovey, Metropolitan Edison Company, from B. J. Snyder, NRC, " Appendix J Exemption Request", September 2,1981.
5.
Final Supplement to the Environmental Impact Statement for Unit 2, NUREG-0112, December 1976.
6.
Fire Protection Program Evaluation, Three Mile Island, Unit 2, June 1977.
7.
Final Programmatic Environmental Impact Statement, NUREG-0683, March 1981.
8.
Letter to Bernard J. Snyder, NRC, from G. K. Hovey, Metropolitan Edison Company, " Technical Specification Change Request No. 26, Addendum B",
LL2-81-0229, October 6, 1981
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