ML20133H577
| ML20133H577 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/24/1985 |
| From: | Cunningham G NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Malsch M NRC OFFICE OF THE GENERAL COUNSEL (OGC) |
| Shared Package | |
| ML20133H476 | List: |
| References | |
| NUDOCS 8508090402 | |
| Download: ML20133H577 (1) | |
Text
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May 24, 1985 Note to: Martin G. Malsch Deputy General Counsel Frcm:
Guy H. Cunningham, III Executive Legal Director
SUBJECT:
UCS LETTERS ON TMI-1 RESTART You have asked me to provide you with sufficient information to respond to the UCS letters to the Commission dated April 18, 1985 and May 3, 1985.
Information regarding the April 18th UCS letter was provided to you on May-17, 1985. The attached Memorandum from Hugh Thompson to me addresses the issues raised by UCS in its May 3rd letter.
Guy H. Cunningham, III Executive Legal Director
Attachment:
As stated cc:
W. Dircks H. Denton DISTRIBUTION:
G. Cunningham chron J. Murray E. Christenbury chron J. Gray J. Goldberg (2)
M. Wagner L. Finkelstein OELD Reading file J. Lieberman J. Gutierrez, RI nd n
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May 24,1985 Docket No. 50-289 MEMORANDUM FOR:
Guy H. Cunningham, III Executive Legal Director FROM:
Hugh L. Thompson, Jr., Director Division of Licensing
SUBJECT:
UCS MAY 3, 1985 LETTER CONCERNING TMI-1 OELD requested DL to provide inputs for a proposed response to a Union of Concerned Scientist letter dated May 3, 1985. Our input is as follows:
The UCS letter states that the staff's presentation to the Commission on April 17, 1985 on the status of TMI-1 in preparation for a restart vote was deficient in three areas involving:
- 1) changes to " lessons learned" items previously certified by the staff to be complete, 2) some conditions of the TMI-1 plant that do not conform to conditions described in the hearing proceeding which were relied upon by the Boards in their decisions, and 3) new Technical Specifications issued by the staff which effectively reverse Board findings that certain equipment is necessary to protect public health and safety by permitting operation of TMI-1 with that equipment inoperable.
Regarding the first and second issue raised by the UCS, the staff has requested the licensee to identify any changes made by the licensee to plant systems and procedures after the staff certified the items as complete and which deviate from commitments documented on the hearing record. We also requested the licensee to state how changes were documented. The licensee in a response dated May 17, 1985 stated that they have procedures which require that any " relevant and material" information be submitted to the NRC and/or Licensing Board as appropriate.
Based-on results of their procedure, plus a polling of appropriate senior management, it is the licensee's judgement that there are no relevant and material changes in plant systems or procedures that have not been reported to the NRC and represent deviations from commitments documented on the hearing record concerning the staff certification items.
Limited independent audit reviews conducted by Region I inspectors indicate there have been numerous changes related to procedures, organization and hardware changes which are in snme way related to the hearing record. None of these changes are considered significant in their effect on the Board decisions. The Region audits support the licensee's conclusion that relevant and material changes have not been made which would have affected the Board decisions. Furthermore, the control of any changes to safety related equipment are covered by the quality assurance program under Appendix B of 10 CFR Part 50.
The licensee is allowed to make changes to equipment without NRC approval if the provisions of 10 CFR 50.59 are satisfied.
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Our responses to the first and second issues are based on an audit of staff, records to. identify any changes, discussions with the licensee, and our evaluation of the effects of such changes. The specific examples cited by UCS for the first two issues were derived principally from staff documents.
The detailed changes and effects of such changes on the previous staff 4
certification are discussed herein.
Issue 1 The UCS contends that some NUREG-0737 lessons learned items previously certified by the staff to be complete are, due to new information, no longer complete.
The specific shcrt term " lessons learned" matters which the Commission determined should be completed prior to restart are grouped within Certification Items 72 through 93 inclusive.3 For the specific examples cited by the UCS, the staff concludes that the changes made by the licensee did not affect compliance with the basis for the certification items found in the hearing records or cause the staff to change its previous certification findings. No new information was provided by the UCS analysis in the examples cited that was not previously considered by the staff. The changes made by the licensee served to enhance plant performance.
Example 1.a pressurizer Safety Valves - Certification Item (CI) 74 This item was certified as complete by the staff based on the licensee's commitment to follow the EPRI test program as described in NUREG-0680.
By letter dated October 28, 1983, GPUN committed to adjust the ring setting of the Dresser Model 31739A code safety valves at TMI-1 so that they would be bounded by the EPRI test results. The Region I Inspection Report 85-03 dated March 19, 1985 noted that revised ring settings were being used by the licensee and they were supported by a technical basis. By letter dated March 4, 1985, GPUN informed NRR of the new ring settings of the code safety valves, their basis for these new settings, and stated that they made these changes under 10 CFR 50.59. The ring settings were chan based upon an analys.is performed by Continuum Dynamics, Inc. (CDI) ged through the B&W Owners Group.
1 SECY-85-64 dated 2/25/85-Three Mile Island, Unit 1 Restart Certification Status n
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' Since the rh g settings are one important consideration in the performance of the safety valves, the staff perfomed a review of the GPUN submittals to determine whether its previous certification conclusion was still valid. The staff Safety Evaluation Report (SER) was issued May 14, 1985.
COI used the valve dynamics simulation code COUPLE, which was validated against EPRI safety valve test data,;to optimize the ring settings of six B&W nuclear plants, including TMI-1, for steam discharge. The staff reviewed the analysis for the new ring settings. The staff concludes that the present ring settings for the TMI-1 safety valves have been shown to produce full lift and stable valve operations for steam discharge conditions by analysis and by comparison with the EPRI test results. The infomation submitted demonstrates the ability of the safety valves to function under expected operating conditions for design basis transients and accidents and conforms to the Appeal Board decision. The staff concluded in BN-85-058 that our previous conclusion regarding the satisfactory completion of Certification Item 74 is unchanged.
Example 1.b NUREG-0737 Item II.F.2 Design and Qualification Criteria for PWR Incore Thermocouples (T/C) Certification Item 155 For incore thermocouples, the UCS questions two actions by the staff as follows:
(1) the staff incorrectly classifies this lessons learned item as complete but it is not fully qualified environmentally, and (2) the staff incorrectly accepted the justification for interim operation until November 1985 for this equipment.
The UCS concern regarding the incore the vocouples is outside the scope of the lesson's learned items which must be complete prior to restart.
It is also outside the scope of the equipment qualification considered by the Comission in CLI-84-11 as necessary for restart and reflected in the scope of Certification Item 155. The UCS concern is related to the 10 CFR 50.49 evaluation of NUREG-0737 Item II.F.2, a long term item that was not required for restart under the Comission's August 9,1979 Order.
In CLI-84-11, the Comission stated that electrical equipment at TMI-1 needed to respond to a TMI-2 type small break LOCA or loss-of-main feedwater accident must be environmentally qualified prior to restart to the radiation levels associated with D0R Guidelines for large break LOCAs.
The staff's Safety Evaluation was issued in a letter to the licensee dated j
April 9,1985. The staff concluded that all the equipment within the i
scope of CLI-84-11, which included the incore thermocouple instrumentation, is environmentally qualified for the radiation level criteria cited in CLI-84-11. The'incore thermocouples are qualified for radiation based on the experience gained from the TMI-2 accident.
4 However, the TMI-1 incore thermocouple cable assembly, which extends from the incore assembly connection to the reactor building penetration h3s not yet been demonstrated to be environmentally qualified in accordance with 10 CFR 50.49 for the combined steam and pressure temperature environment associated with a large break LOCA or steam line break accident. The staff's letter to GPU dated March 29, 1985 granted the licensee's request for an extension of the March 31, 1985 environmental qualification deadline and found the justification for interim operation (JIO) sufficient to support operation during the requested extension period.
Notwithstanding that the 10 CFR 50.49 evaluation is not a certification matter, the staff does not agree with UCS that GPU has not provided a valid JIO for operation.of TMI-1 without environmentally qualified incore thermocouple instrumentation. The staff's basis for its evaluation of the JIO is described herein.
-The incore thermocouples are used to determine the margin to subcooling when the reactor coolant pumps are not running. The margin to subcooling is used procedurally to determine when high pressure injection can be throttled.
For a large break loss-of-coolant accident (LOCA) and for a large small break LOCA (say greater than 0.085 ftri, if the thermocouples failed and the reactor coolant pumps are not running, the operator is prevented by procedure from throttling high pressure injection until low pressure injection (LPI) flow is established at 1000 gpm. Once LPI is established, core temperatures can be measured utilizing instrumentation in the LPI system which is fully qualified.
Preventing HPI from being throttled under these circumstances will insure that proper conditions will be established for LPI in less than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.
For a small break LOCA (say less than 0.085 fte), the resulting containment environment is such that thermocouple cable / connector failure is not expected based upon TMI-2 experience.
Accordingly the JIO for the incore thermocouple cable / connectors is acceptable based on administrative controls through plant procedures, use of alternate qualified equipment, and partial test data from TMI-2.
Issue 2.
The UCS contends that some conditions' of the TMI-1 plant do not conform with the conditions described in sworn testimony during the restart proceeding which were relied upon by the Licensing Board and Appeal Board in deciding contested issues in favor of restart.
As for Item 1, the short-term lessons learned certification items, the staff similarly concludes from its audit that changes noted herein from plant conditions described in the hearing record did not affect compliance with the basis for the certification items found in the hearing records or cause the staff to change its previous certification finding.
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Example 2..
Emergency Power to Pressurizer Heaters.
Certification Items 72, 110, and 111 For the UC5's example of the emergency power supply to the pressurizer heaters, a Regional Inspector reviewed the licensee's actions and the detailed records available at the site. The inspector concluded that the licensee's actions were proper and closed the issue. Nevertheless, since the issue was brought up by the UCS, the staff has reviewed it in detail with the following results:
The electric power supply to the pressurizer heaters at TMI-1 was modified to include the capability to supply power to the pressurizer heaters from the Class 1E buses in the event of loss of offsite power. The Class IE buses are protected by safety grade circuit breakers to protect against failure in the pressurizer heaters which are "non-safety grade." The design also includes Class IE undervoltage relays to trip Class IE circuit breakers to protect the Class IE buses from faults in the pressurizer heater circuitry. The licensee.
had earlier proposed preliminary settings for the undervoltage relays of 430 volts and 1.5 second time delay.
Subsequently, the licensee utilized new settings of 420 volts and 1 second time delay. The UCS had questioned the new settings of undervoltage relays. UCS concerns are:
(1) the pressurizer heaters will be connected to the Class IE bus without first stabilizing the load; and (2) the new settings will not protect the Class IE bus from a fault in the pressurizer heaters.
The pressurizer heaters will be connected to the Class 1E bus only after the automatically connected loads have been sequenced on the Class 1E bus and the voltage has stabilized after an accident or loss of offsite power. This has been reconfimed by the licensee in recent discussions. The licensee indicates that the change in settings of the undervoltage relays was made to pemit subsequent manual connections of loads as necessary. The original settings on the undervoltage relays were preliminary and did not account for the subs'equent manual connection of loads. The purpose of the undervoltage relays is to act as protective devices to backup the circuit breaker overcurrent trip devices. Based upon review of licensee information, the new settings (420V and I second) provide' adequate protection for faults in the pressurizer heaters as well as allowing the manual connection of other loads as necessary without tripping the pressurizer heaters.
Based upon our discussions with the licensee and review of licensee information concerning the emergency power supply for the pressurizer beaters, the staff has concluded that (1) the pressurizer heaters will be manually connected to the Class IE bus only(after the automatically connected loads on the Class IE bus have stabilized, 2) the new setting (420 volts and 1 second) is adequate to protect and isolate the Class IE bus from faults in the pressurizer heaters, and (3) these changes will not degrade the Class 1E bus.
Based upon the above conclusion, there is no safety concern regarding the emergency power supply to the pressurizer heaters.
In addition, a license
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condition requires that before the pressurizer heaters are connected to the emergency power supply at TMI-1, the reactor shall be subtritical or in hot standby condition. Therefore, this modification does not change our previous position that Certification Items 72,110, and 111 are complete, Issue 3.
The last contention by the UCS is that many of the new TMI-I Technical Specifications proposed by GPU and approved by the staff effectively reverse Licensing Board and Appeal Board findings that certain equipment is necessary to protect the public health and safety by permitting operation of TMI-1 with that equipment inoperable.
Four examples were provided by the UCS as follows:
(1) subcooling margin instrumentation (2) position indication for the Pressurizer PORV and safety valves, (3) emergency feedwater flow instruments, and (4) reactor coolant system high point vents. Three of these Technical Specifications were issued via Amendment No. 78 to the license dated October 20, 1982. The Technical Specifications involving high point vents was issued as Amendment No. 97 dated June 21, 1984. Copies of both of these amendment packages were provided directly to the Hearing Boards.
We examined board findings concerning restart and were unable to locate any board statements that the above equipment is necessary to protect the health and safety of the public nor that the Board had addressed the matter of Technical Specifications on these items. The equipment was recomended by the Lessons Learned Task Force in NUREG-0578 and implemented by NUREG-0737.
In particular the following TMI action items in NUREG-0737 relate to this equipment.
(1) Subcooling Margin Monitor II.F.2 (2) Valve Position Indication II.D.3 (3) EFW Flow Indication II.E.1.2 (4) High Point Vents II.B.1 Technical Specifications were required for each of these as stated in Table C.1 of NUREG-0737 and have been incorporated in the Technical Specifications of TMI-1.
In contrast to the statements of UCS, the staff concludes that the Technical Specification requirements for the valve position indication and high point vents are adequate to protect the health and safety of the public.
In the case of the EFW flow indication and the subcooling margin monitor, the staff concludes that the operability requirements in the present Technical Specifications are not consistent with the Standard Technical Specifications.
The more restrictive requirements of the latter are currently being considered by the staff for generic backfit to several B&W plants, including TMI-1. The staff concludes, however, that the present Technical Specifications do not represent a public health or safety concern that would justify imediately effective enforcement action.
~ 7-The specific' details of our conclusions are as follows:
a.
Subcooling Margin Instrumentation
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This issue was previously addressed in our response to an April 5,1985 UCS letter. The present Technical Specifications for TMI-I permit continued plant operation with only one of the two channels of the subcooling margin monitors operable; corrective actions are required only when both channels become inoperable. Since the subcooling margin meters perfonn no automatic safety function and since a backup method is available using core exit thermocouples and steam tables when the subcooling margin meters are inoperable, the staff found that plant operation is justified up to seven days, which is a limiting condition of operation (LCO) in the Technical Specifications.
Surveillance requirements of Technical Specifications include a check each shift and a monthly test when the plant is hot, i.e., T average is greater than 525'F. The surveillance requirements are consistent with those required for safety-related equipment.
However, the operability requirements of the present Technical Specifications are not consistent with the Standard Technical Specifications which require that both channels be operable at all times, and that corrective action be taken within seven days if any one of the two channels should become inoperable.
If both channels should become inoperable, corrective action would be required within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The interim Technical Specifications now in place at TMI-1 will be considered for upgrading when the staff completes its evaluation of the design criteria of inadequate core cooling (ICC) instrumentation per item II.F.2 of NUREG-0737.
Based on the staff's evaluation of the present Technical Specifications discussed above, the staff concludes that pending on our ICC instrumentation review, the present Technical Specification does not represent a public health or safety concern that would justify imediately effective enforcement action.
- b. Valve Position Indication (TMI Action Plan Item II.D.3)
Positive indication of reactor coolant relief and safety valves is required in NUREG-0737 "so that appropriate operator actions can be taken." Valve position indication does not have to be safety grade if " reliable single-channel direct indication" is 'provided and that " backup methods of determining valve position are available and are discussed in the emergency procedures as an aid to operator diagnosis of an action."
In the event that one of the two valve position indicators was out of service and for some reason, the remaining one failed at the same time a safety or relief valve stuck open, there are a number of diverse protective measures to ensure the event would be successfully mitigated.
First, pressure relief tank pressures would be expected to be available, and c1rrent operator training would allow the operators to properly interpret this pressure to infer a stuck-open PORV.
In addition the operators are Hstructed to close the PORV block valves in the event the symptoms of a primary system LOCA are observed.
Finally, revised operating procedures would instruct the operator not to terminate HPI unless conditions existed in the plant that assured the core was adequately cooled and decay heat was being removed.
i Because of the multiple means of protecting the plant against a stuck-open._ _,
PORV or safety valve, coupled with the fact that the valve position indication does not provide any direct mitigative function, nor does it i
ensure correct operator action, the current TMI-1 Technical Specifications are acceptable and comply with the intent of TMI Action Plan Item II.D.3.
- c. Auxiliary Feedwater System Flow Pate Indication (TMI Action Plan Item II.E.1.2 Part 2)
Section 11.D11.2 Part 2 of NUREG-0737 (clarification of TMI Action Plan Requirements), requires that " safety-grade indication of auxiliary feedwater flow to each steam generator be provided in the control room for B&W plants.
The intent of this requirement is to assure that a reliable indication is provided in the control room from which the operator (s) can ascertain the actual performance of the auxiliary feedwater system (AFWS) when called upon to perform its safety function. This is consistent with the requirements of General Design Criterion 13 (Instrumentation and Control).
NUREG-0737 requires that AFWS flowrate indiation for B&W plants for each steam generator satisfy the single failure criterion (i.e., a minimum of two flowrate indications), be seismically and environmentally qualified, have the capability for testing during plant operation, and conform to staff requirements concerning channel independence and control / protection system interaction. NUREG-0737 makes a distinction in this regard between B&W plants and other PWRs; other PWRs have the option of providing a minimum of two flowrate indications or one flowrate indication and one steam generator level indication. This distinction is due to the fact that, for the once-through steam generators (OTSG) used in B&W plants, steam generator level does not provide a true indication of heat removal capability.
In the OTSGs, the feedwater is sprayed on the tubes.
It is not necessarily indictive of inadquate heat removal or feedwater flowrate if the indication level is very low or zero. Also, AFW pump discharge pressure indication is not an acceptable substitute for flowrate because it does not provide a direct, or a positive indirect, measurement of the desired parameter.
Model Technical Specifications (TS), intended to assure that facility operation is maintained within the limits determined acceptable following implementation of TMI Action Plan requirements, would require that both AFWS
~ flowrate indication channels for each steam generator be operable during operation (i.e., Modes 1, 2 A 3; power operation, startup, and hot standby respectively).
If one of the two channels becomes inoperable, the corresponding action statement would require that it be restored to an operable status within seven days, or that the reactor be placed in a hot shutdown condition within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
If both channels become inoperable, one of the channels must be restored to an operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, or the reactor placed in a hot shutdown condition within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
-9 The current.TMI-1 Accident Monitoring Instrumentation TS (Section 3.5.5) only requires one AFWS flowrate indication channel per steam generator to be operable in Modes 1, 2 & 3.
If neither AFWS flowrate indication channel is operable, one of the inoperable channels must be restored to an operable status within seven days, or the reactor be placed in hot shutdown within the next six hours and in cold shutdown within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. These TS are inconsistent with the model TS issued by the staff for implementation of TMI Action Plan requirements, the proposed TSs for RG 1.97 instrumentaiton, and the intent of both RG 1.97 and NUREG-0737 (Item II.E.1.2 Part 2) regarding safety related indication of AFWS flow for B&W plants.
The more restrictive requirements for AFWS flow indication of the model Technical Specifications are currently being considered for generic backfit to several B&W plants, including TMI-1. The present Technical Specifications at TMI-1 requ re that the three EFW pumps and associated flow paths shall be i
operable when the reactor coolant temperature (RCT) is greater than 250*F. With one pump or flow path inoperable, the licensee has 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> te take corretive action or be in cold shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. With more than one pump or flow path inoperable, the licenee must restore the inoperable pumps or flow paths to operable states within one hour or be in cold shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Therefore, the staff concludes that the present TMI-1 Technical Specifications for EFW flow indication do not represent a public health or safety concern that would justify immediately effective enforcement action.
- d. High Point Vents (TMI Action Item II.B.1)
High Point Vents are required in NUREG-0737 "to vent non-condensible gases from the RCS which may inhibit core cooling during natural circulation." The requirement for high point vents is further implemented by 10 CFR 50.44 which requires the vents to be provided for removal of non-condensible gas which might cause loss of core cooling following a loss-of-coolant accident. The staff evaluated the effect of non-condensible gas, which might be generated within the reactor system following a design basis small break LOCA in NUREG-0565. The effect of the expected gas release would be insignificant.
The high point vents are not required to mitigate any design basis accident and are only required to relieve non-condensible gas as might be released from a beyond design basis event. The vents are, therefore, defense in depth equipment.
There are four vents in the TMI-1 system: one inteach loop, one on the pressurizer and one on the reactor vessel head. The current Technical Specifications permit operation of TMI-1 with one loop vent inoperable at all tines and permit continued operation for 30 days following failure of another vent.
If a third vent becomes inoperable, the licensee is required to take corrective action to restore at least two of the inoperable vents within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in hot shutdown within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The occurrence of an accident for which use of high point vents would be required for natural circulation restoration would be extremely unlikely and would involve failure of redundant safeguard 1
equipment apd loss of offsite power so that forced flow would not be availiible.
For the high point vents not to be available would require that.{a) the loop A vent was out of service, (b) a degraded core accident which results in a damaged but recoverable core in which significant amounts of non-condensible gases were generated, and (c) a loss of power to the remaining vent train.
Moreover, even with this scenario, it is questionable whether additional core damage would actually occur. Because of the extremely low probability of this scer.ario, we do not believe the current Technical Specifications for high point vents pose an unacceptable risk to the health and safety of the public and are, therefore, acceptable.
The UCS then requested the Commission to direct the staff to take three actions. The first two actions involve recertification of all items previously certified as complete. Based on the examples of changes made to date as discussed earlier, we do not believe there is a need to recertify the issues.
However, the staff's final certification of all the required actions will be provided to the Commission by May 29, 1985, and such certification reflects the staff's judgement that the concerns raised by UCS do not effect the certification of any item. The third action requested was that the entire Technical Specifications be reviewed by both the staff and the parties and amended as necessary to require, as limiting conditions for operation, the operability of all equipment relied upon by the Boards. The staff reviewed 5
and approved the present Technical Specifications for TMI-1 and believes that they do correctly reflect the board's decisions. The UCS has provided nn examples of where the Technical Specifications reverse Board findings. There is therefore no basis for a complete review of all Technical Specifications by all parties.
Hu
. Thompson, r.
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