ML20054C394
| ML20054C394 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/09/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20054C389 | List: |
| References | |
| NUDOCS 8204200509 | |
| Download: ML20054C394 (7) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION GPU NUCLEAR CORPORATION METROPOLITAN EDIS0N COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY DOCKET NO. 50-320 i
THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 2 Introduction By letters dated February 23,1981 (Reference 1), March 18,1981 (Reference 2),
and October 6, 1981 (Reference 8), the licensee proposed changes to the Recovery
~
mode Technical Specifications, Appendix A/B for Three Mile Island Unit'2 (TMI-2).
Because of the variety of types of changes requested, each has been separated and addressed individually. The requested changes to Appendix B were issued on May 6,1981 in license Amendment No.15; the requested changes to Appendix A are discussed herein.
Evaluation (1)
Instrumentation A.
The Reactor coolant flow indicators RC-14A-dPT1, RC-14A-dPT2, RC-14A-dPT3, RC-14A-dPT4, RC-148-dPT1, RC-14B-dPT2, RC-14B-dPT3, and RC-14B-dPT4 are differen-tial pressure transmitters located in hot legs A&B of the Reactor Coolant System.
4 They are currently required to be operable in Table 3.3-9 of the proposed Technical Specifications. With the reactor coolant pumps operating, the pressure drop across a Gentile tube for each leg is measured and converted to a corresponding flowrate. The differential pressure /flowrate signal is then transmitted to the Reactor Protection System for reactor trip signal generation. This signal genera-tion is presently not necessary since in the Order for Modification of License dated July 20, 1979, the licensee's authority was limited to maintenance of the facility in the present shutdown mode.
The reactor coolant pumps which could produce the minimum required flow needed to utilize the indicators have not been i
used since the last pump was stopped on April 27, 1979 and natural circulation of the primary system was initiated.
There is no present requirement or future need for using the reactor coolant pumps for core cooling during the remainder of the Recovery Mode. Therefore, the requirement to maintain these pumps in an operable
,gg status was deleted by an Amendment of Order dated November 14, 1980.
O M G.
i 88 The licensee has proposed to modify Table 3.3-9 of the proposed Technical gg Specifications by deleting the operability requirement for all of the differential pressure transmitters used for reactor coolant system flow indication.
Deletion o
8:c of'the associated Surveillance Requirement in Table 4.3-6 has also been requested.
$8 The staff finds that the present instrumentation used to monitor the shutdown oc condition of the reactor such as pressure and temperature transmitters, reactor U
coolant system boron sampling, and two operable source range neutron monitoring 85 instruments adequately monitors the system and therefore finds the proposed changes N-acceptable.
. B.
The Reactor Building Emergency Spray and Core Flooding Systems were deleted from the proposed Technical Specifications in the February 11, 1980 Order.
The licensee has therefore requested to delete from Table 3.3-10 of the proposed Technical Specifications the operability requirements of the Core Flood Tank Level and Reactor Building Spray Pump Flow Instrumentation and the associated Surveillance Requirements in Table 4.3.10.
Based on the above discut eion, we find that these instruments are not required to monitor the reactor i nlant system in its present condition and therefore find the proposed changt.s acceptable.
(2)
Hydrogen Purge Cleanup System The February 11, 1980 Order 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 with established off-site dose limits.
The purge began on June 28, 1980 and was completed on July 11, 1980.
Subsequent permission to perform 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 therefore requested that the Hydrogen Purge Cleanup System, Section 3.6.4.3, be deleted from the proposed Technical Specifications and Surveillance Require-ment 4.6.4.3.
The staff has noted that the only credible event that could produce substantial amounts of hydrogen would be a recriticality accident concurrent with a temperature increase severe enough to cause the zircaloy cladding to begin decomposition.
Recriticality was discussed in the Final Programmatic Environmental Impact Statement (PEIS) for TMI-2 issued in March 1981.
Paragraph 4.1 of the PEIS states that "the most probable (although very unlikely) cause of recriticality was found to be boron dilution, which would be a slow enough process that any approach to criti-t cality can be detected and remedied."
This statement is still valid; there-l fore, the staff has concluded that this accident need not be designed against in referente to hydrogen production.
Since the Hydrogen Purge System's purpose was fulfilled upon the completion of the June 28, 1980 purge and the use of the system will not be required to maintain the containment building in a safe l
condition for the remainder of the Recovery Mode, the Staff finds that the deletion of Sections 3.6.4.3 and 4.6.4.3 is acceptable.
(3)
Fire Protection A.
Proposed Technical Specifications 3.7.10.2 d and e currently require 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.
The reason for this requirement was to minimize the fire hazards associated with the charcoal in the filter train.
The primary purpose for this charcoal was to remove gaseous iodine (I-131) from the effluent prior to release to the atmosphere.
However, the iodine-131 levels in the containment building have been determined to be minimal.
Therefore the charcoal has since been re-moved from both exhaust filters.
With this removal, the fire hazard associated with these filters has also been eliminated.
The licensee has therefore proposed l
.. that the requirement to maintain the associated deluge systems in an automatic initiation status be deleted.
The deluge system would be isolated, thereby placing it in a manual mode rather than the previously required automatic mode.
The staff agrees with Met-Ed that the manual mode would prevent spurious signals from needlessly activating the deluge system and since the fire detectors asso-ciated with this system are not affected, an alarm would still be provided in the event of fire in the filter area.
Therefore, the staff approves the requested modification to proposed Technical Specifications 3.7.10.2 d and e.
B.
The licensee has requested the deletion of the requirements reflected in proposed Technical Specification 3.3.3.8 which requires the fire detection instru-mentation for the Balance of Plant Diesel to be operable as imposed in the February 11, 1980 Order.
By letter dated April 28, 1980, the licensee proposed to delete the operability requirements for the B0P diesel generators and the 13.2Kv transmission line, replacing their capabilities with an existing 230 Kv grid system.
This modification was approved in the August 11, 1960 Modification of Order.
In that approval, the Safety Evaluation determined that the 230 Kv grid system adequately provided access to back up power capability for all of the plant equipment essential to the preferred reactor cooling modes.
Since the diesels are located outside of the restricted area fence, are not located near any vital equipment, and will no longer be required to perfonn a safety function, the staff concurs with the licensee's proposal.
C.
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 suppres-sion 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 (Reference 6). The Bases for the Proposed Technical Specifications also reflects the same conclusion.
Therefore, the staff has concluded that enough conservatism is present in the proposed Technical Specifi-cations 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 witnout impairing the Fire Suppression System.
(4)
Containment Systems A.
Section 3.6.1.1 of the Proposed Technical Specifications presently contains a grammatical contradiction by requiring that with one containment isolation valve per containment penetration open or inoperable, maintain the affected penetration (s) closed, with either action (a) and action (b) taking place with no alternative de-signated. The staff agrees with the licensee that action (a) should be followed by "or" which will indicate two action alternatives as intended in this requirement.
B.
Proposed Technical Specification 3.6.1.3 (b) presently requires the con-tainment air locks to be operable with an overall leakage rate of less than or equal to 0.05La at Pa, 56.2 psig where La is the maximum allowable leakage rate (%/24 hrs) and Pa is the calculated peak containment internal pressure.
The leakage requirements are set forth pursuant to the requirements of Title 10, Part 50, of the Code of Federal Regulations, Appendix J.
These airlocks are installed ae an integral part of the containment structure providing access to the reactor bui;iing while main-taining a barrier against the possible release of airborne contamination of the environment.
,, The licensee has requested a modification of the leak rate criteria (Proposed Technical Specification 3.6.1.3.b) because verification of compliance cannot be made without unacceptably high radiation exposures to personnel.
In addition, the June 28, 1980 purge significantly decreased the airborne contami-nation levels, greatly reducing the concern over containment atmosphere leakage.
Based in part on the above philosophy, the licensee has previously been granted an exemption from certain requirements of Appendix J upon which requirement 3.6.1.3(b)isbased.
Therefore the proposal by the licensee to delete Section 3.6.1.3 (b) from the proposed Technical Specification and delete Section 4.6.1.3 (b) from the surveillance requirements is approved.
However, Section 3.6.1.3 (a) has not been deleted.
The footnote to surveillance requirement 4.6.1.3 (a) indicating that item (a) is an exemption to Appendix J has been removed.
Since Appendix J was recently modified (45FR62789) and now requires the same leak rate testing on airlock doors as is stated the (*) and reference statement is not longer requit ed.
C.
Proposed Technical Specification 5.2.2 presently requires that the contain-ment building be designed and maintained for a maximum internal pressure of 60 psig and a temperature of 286 0F.
The licensee performed an analysis evaluating the peak containment pressure under accident conditions in support of a request for exemption from Appendix J (Reference 3).
This analysis, which was independently verified by the NRC staff, concluded that the maximum potential containment building pressure was 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 illustrated that even with the failure of penetration R-626, the offsite release would be less than 400 uCi 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 assessed in the
" Final Programmatic Environmental Impact Statement Related to the Decontamination and Disposal of Radioactive Wastes Resulting from the March 28, 1979, Accident at TMI-2", dated March 1981. Also to minimize the potential for a fire induced over-pressure, the licensee has instituted fire hazards controls for the containment and is currently rewriting applicable procedures.
In addition, piping and electri-cal penetrations other than R-626 are being re-designed to 5 psig.
R-626 will remain as is until accessbility and personnel exposure permit its possible upgrading.
The staff's independent evaluation agrees with the licensee's con-clusions and, therefore, Section 5.2.2 has been modified to state a containment design pressure of 2 psig.
The design temperature will remain at 286 0F.
A statement has also been added to Section 5.2.2 at the licensee's request, which l
states that any maintenance to be perfonned on the containment building to main-I tain this design pressure shall be done per occupational exposure considerations.
1 D.
The maximum containment pressure was limited to <0 psig in proposed
., Technical Specification 3.6.1.4 by the issuance of the February 11, 1980 Order.
At that time airborne contamination levels were high and the required negative pressure value insured that all leakage would be into the building and not out.
Since the Order was issued, the Reactor Building has been successfully purged.
This decreased the airborne contamination levels to a value that normal leakage will not affect the health and safety of the public.
Based on this discussion, the licensee has requested a modification of proposed Technical Specification 3.6.1.4 by increasing the internal pressure limitation to + 1 psig.
In eval-uating this proposal the staff agrees that the present airborne contamination levels inside containment are presently low. However, future operations that are expected to take place, such as containment building decontamination and the eventual removal of the reactor vessel head, have the potential for increas-ing the airborne radioactivity levels.
With this increase, an outleakage from the building resulting from an i'nternal positive pressure would be undesirable.
The staff is of the opinion that because of this potential, the request to increas the pressure to a maximum of + 1 psig in proposed Technical Specifi-cation 3.6.1.4 is unacceptable and furthermore, unnecessary since the Reactor Building purge system can readily maintain a negative internal pressure by exhausting through HEPA filters.
Exhaust flow through the HEPA filters will minimize the release of radioactive particles to the environment.
This opinion was verbally transmitted to the licensee and subsequently the licensee has with-drawn the originally requested change per LL2-81-091, Barton to Snyder, dated December 4, 1981.
(5)
Control Room Emergency Air Cleanup System The proposed Technical Specifications do not require that the control room air inlet radiation monitor remain operable for all accident conditions.
Presently, a failure of the control room air inlet radiation monitor could prevent the auto-matic shif ting of the control room emergency air cleanup system to the recircu-lation made 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 l
The staff agrees that this change enhances the safety of control room personnel and therefore concurs with the modification.
l (6)
Personnel Qualifications In accordance with the Proposed Technical Specifications for TMI-2, the plant Operations Review Committee (PORC) is composed of a Chairman, one member who meets or exceeds the qualifications of Regulatory Guide 1.8, September 1975, and seven j
members who meet or exceed the qualification requirements of Section 4.4 of I
In this change request, the licensee proposes to change the qualification requirements of the seven members from those specified in Section 4.4 of ANSI N18.1-1971 to those specified in Section 4.7.2 of ANSI /ANS 3.1-1978, which is a later version of ANSI N18.1-1971.
l Section 4.4 of ANSI N18.1-1971 describes the qualification requirements of l
" Professional-Technical" personnel.
Of these personnel, the responsible person in reactor engineering or physics was to have a bachelor's degree in engineering or the physical sciences and at least two years of experience in such areas as l
reactor physics, core measurements, core heat transfer, and core physics testing program.
Other Professional-Technical personnel (in the areas of instrumenta-tion and control, radiochemistry, and radiation protection) were to have at least five years of experience, of which two years should be related technical l
1
.. training.
In each case, for these other Professional-Technical personnel, up to four years of the five years of experience could be fulfilled by related technical or academic training.
Section 4.7.2 of. ANSI / ANS 3.1-1978 describes,the qualifications of Staff Specialists who perform independent reviews of operational phase activities at nuclear power plants.
These individuals are to have a bachelor's degree in engineering or the physical sciences as appropriate and three years ofIn professional level experience in their respective fields of specialty.
special cases, eight years of experience in the specialty field may be acceptable without a degree. Also, credit toward experience may be given for advanced degrees in any of the specialized fields on a one-for-one basis up to a maximum of two years. The specialty fields covered are administrative control, nuclear power plant operations, nuclear engineering, metallurgy, quality assurance, non-destructive testing, chemistry and radiochemistry, instrument and controls, radiological safety, and mechanical and electrical engineering.
The proposed change to the qualifications of the PORC members thus would substitute 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.
Thus, we find that the proposed qualifications for the PORC members are higher than the qualifications required by the existing Technical Specifications and, accordingly, we conclude that the proposed change to the Technical Specifi-cations is acceptable.
(7)
Fuel Handling Building / Auxiliary Building Air Cleanup Systems The " action" statement of Proposed Technical Specification Section 3.9.12 (a) states that "with the fuel handling buildina/ 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 system 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 Therefore, the staff concurs with a modification to other operating (system.b) as requested by the licensee to eliminate ambiguity.
3.9.12 (a) and Environmental Considerations Based on the above evaluations, the approved changes in the proposed Technical Specifications will not result in any environmental impact beyond those considered in the Final Programmatic Environmental Impact Statement, NUREG-0683 (Reference 7) and the Final Supplement to the Environmental Impact Statement for Unit 2, NUREG-0112 (Reference 5).
The staff has detemined that these changes to the proposed Technical Specifications do 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.
Having made this determination, the staff has furtner concluded that this modification of the proposed Technical Specifications involves an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 51.5 (d) (4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection
9 with the issuance of this proposed Technical Specification change.
(8)
Standby Pressure Control System Paragraph 4.1.1.1.j.of the Surveillance Requirements presently indicate 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 by using more accurate system tenninology what tanks are to be sampled. 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 a fdotnote 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.
Conclusion Based upon the st'aff's review of the proposed modifications to the proposed Technical Specifications, the staff finds that the licensee's changes are accept-able with the exception of modifications that were withdrawn by the licensee as discussed in part 4(D) of this safety evaluation.
Based on the review of the licensee's approved requests, the staff has concluded that (1) the modifications do not authorize a significant change in the plant's operation; (2) the modifi-cations do not involve a significant increase in the probability or consequences of accidents previously considered or a significant reduction in a margin of safety and therefore, does not involve a significant hazards consideration, (3) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the modified manner, and (4) such activities will be conducted in compliance with the Commission's regulations and the issuance of this Amendment of Order will not be inimical to the comon defense and security or to the health and safety of the public.
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