ML19305B878
| ML19305B878 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/17/1980 |
| From: | Swartz L NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Sholly S AFFILIATION NOT ASSIGNED |
| References | |
| NUDOCS 8003200519 | |
| Download: ML19305B878 (21) | |
Text
TEWA UtilTED STATES OF AMERICA fiUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICErJSING BOARD In the Matter of
)
)
METROPOLITAN EDISON COMPANY
)
Docket flo. 50-289
)
(Three Mile Island, Unit 1)
)
fiRC STAFF RESPONSE TO THE FIRST SET OF INTERROGATORIES SUBMITTED BY INTERVEN0R STEVEN C. SHOLLY On February 29, 1980, the fiRC Staff responded to most of Steven C. St.olly's first set of interrogatories to the NRC Staff dated January 17, 1980. All but four of the remainder of those interrogatories are answered in this pleading.
Each interrogatory is restated and a response provided.
Where appropriate, the NRC Staff has invoked that portion of the Cormission's Order of August 9, 1979 (Slip 0. at 11) which allows as an adequate response to a discovery 2
request a statement that information is available in the Local Public Docunent Rooms and guidance as to where the information can be found.
Following the responses to the interrogatories are affidavits identifying the individuals who prepared the responses and verifying them.
8003200 5/9
Interrogatory 2-2 Question Has staff r~eceived from the licensee details concerning the qualification testing program for the pressurizer PORV, conduct of which is required by a recommendation from the Lessons Learned Task Force?
If so, identify documents relating to this matter.
If not, will this qualification testing program be required to be complete prior to restart? If not, why not?
Response
Metropolitan Edison Company (Met-Ed) in the TMI-1 Restart Report has referenced a test program currently being developed by EPRI/NSAC which is to meet the long-term requirement on PORV testing as set forth in Section 2.1.2 of NUREG-0578.
In the staff's SER on TMI-I restart the staff requires that Met-Ed justify that the EPRI test program is applicable to the TMI-1 PORV.
The staff considers it acceptable to have PORV testing completed in the long term since even if a PORV should fail open (a) sensors to be installed at the PORV discharge will allow the operator to detennine if the PORV is open or shut; (b) B&W LOCA guidelines require the PORV block valve to be closed early in a LOCA; (c) improved power supplies are required by Section 2.1.1 of NUREG-0578 for the PORV and its associated block valve; and (d) small break LOCA emergency procedures are to be upgraded at TMI-1 to meet the staff approved B&W guidelines The following documents were used in preparing this response: NUREG-0578; TMI-1 Restart Program; Commission Order concerning Met-Ed, dated August 9, 1979; and IE Bulletins79-05A and 79-05B.
Interrogatory 3-4 Question Is it staff's position that Unit I complies with GDC-35 with regard to metal-water reactions following a LOCA? Explain your answer.
Response
Acceptance criteria for metal-water reactions are defined in 10 CFR 50.46,
" Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear power plants," in the sections on maximum cladding oxidation and maximum hydrogen generation.
The :nalysis that has been pcrfctmed for a spectrum of s all trea'< locaticas
..ich dmrs'. ates that th2 specific part. meters of 10 CFR 50.?6 will not be aceeded are those which have becn previously suhaitted in res;cnse to a Scimber 27, 1974 Comission order. These analyses '..cre found in complience by the staff in a letter from R. W. Reid (NRC) to Matropolitan Edison dated May 18, 1976.
On April 27, 1978, as the result of an analysis deficiency involving a particular break location, the ECCS analyses for TMI-l were not found in conformance with 10 CFR 50.46 requirements. The NRC granted Metropolitan Edison an exemption to 10 CFR 50.46 on April 27, 1978, but limited operation to 91 percent of design full power.
Based on additional information submitted by Metropolitan Edison on May 3,1978, the Staff removed the 91 percent power restriction and allowed full power operation.
However, we did not conclude that the ECCS analyses conformed to 10 CFR 50.46 requirements. Theexemptionwaslimitedtoaperiodoftimenecessarytocomplet$
and review the revised calculations.
The revised calculations have not yet been received by the Staff.
These will be required before restart will be allowed.
Since the exen.pticn was grcated, the staff has id ntified an additional small break scenario considered within the design basis of the plant which would lead to consequences in violation of 10 CFR 50.46 limits.
This involves a small break LOCA in which the reactor coolant punps are post ulated to experience a delayed trip.
As a result, NRC issu2d 3uilatin 79-05C requiring that all coolant pumps be tripped upon reactor trip cr.d liPI acteation on low pressure.
In f:Sy of 1979 BEN submitted a generic evaluation of small reactor c:olant system breaks.
This evaluation was performed to demonstrate that the s::.all break spectrum analyzed in the FSAR to show conforrance to 10 CFR 50.46 was still acceptable, and to provide information on
?SW plant r:sponse to small breaks under conditions not normally ccnsid: red for FSI,R an: lyses.
These incleda very small breaks, which, while relying on natural circu-lation for decay heat removal, do not produce core uncovery and therefore, do not challenge 10 CFR 50.46 limits. Also analyzed were loss of all feede:ater cvants for the purpose of identifying available times for c:rrectivs opcrator actions, and asymetric auxiliary feedwater (feed only one st:5m ger.erator).
From ticse analyses in the May,1979 submittal, it was concluded that cultiple failures must be assum.2d before accident scenarios not identified in the FSAR could be postulated which would lead to core uncovery and challenge 10 CFR 50.46 limits.
ETSB INPUT FOR RESPONSE TO INTERROGATORIES FROM S. C. SHOLLY 'i TMI-l RESTART 5 According to NUREG-0578 at page A-37, "A recent survey of existing gaseous effluent monitoring capabilities of operating plants shows that less than 20 percent of operating plants have monitors that would have stayed on scale under the conditions of the TMI accident.
It can also be shown, however, that the potential releases from postulated accidents may be several orders of magnitude higher than was encount-ered at TMI. Under such circumstances, none of the effluent monitors now in service at any operating plant would remain on scale." With regards to this quotation from the Lessons Learned Task Force, answer the following questions:
A.
What is total amount of radiation released during the Unit 2 accident?
Response
Based on "TMI, A Report to the Commissioners and to the Public" by the NPC Special Inquiry Group (Rogovin Report), we estimate the quantity of noble gas radionuclides released during the TMI-2 accident, for the period March 28, 1979 to April 3,1979, to have been approximately 6
2.5 x 10 Curies.
The quantity is in good agreement with 2.37 x 106 Curies estimated by J. A. Auxier, et al, " Report of the Task Group on Health Physics and Dosimetry to President's Comnission on the Accident of Three Mile Island," October 1979 and the early estimates of about 6
10 x 10 Curies reported in NUREG-0600, " Investigation Into the March 28, 1979 Three Mile Islend Accident by Office of Inspection and Enforce-ment," August 1979, by Met. Ed., " Third Interim Report-July 16, 1979 and by Pickard, Lowe and Garrick, TDR-TM1-ll6, July 31,1979.
For i
radioiodines, the calculated quantity released, based on samples collected from the ventilation system charcoal adsorbers and reported in "TMI, A Report to the Commissioners and to the Public," for the per-iod March 28, 1979 to April 3,1979 was approximately 6.5 Curies mea-sured as iodine-i31 and in the extend period April 4,1979 to May 8, 1979, there were additional releases of approximately 8 Curies of iodine-131, as reported by Pickard, lowe and Garrick, TDR-TM1-ll6, July 31,1979.
B.
What is the total amount of radiation released in the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the Unit 2 accident?
Res ponse Based on fiUREG-0600, we estimate the quantity of noble gas radionuclides 6
released during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> was approximately 1.5 x 10 Curies.
Based on the Rogovin Report, we estimate the quantity of radiciodines released during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> was approximately 0.95 Curies, mea-sured as iodine-131.
C.
What is the maximum release rate (in Ci/sec) during the Unit 2 accident?
Response
From information given in the Rogovin Report and TDR-TMl-ll6, we esti-mate the maximum release rate of noble gas radionuclides to have occurred between 0745 hours0.00862 days <br />0.207 hours <br />0.00123 weeks <br />2.834725e-4 months <br /> on fiarch 28 (when the vent monitor HP-R-219 went off-scale) and 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> on fiarch 29, 1979 at approximately 35 Ci/sec (Xe-133).
From the Rogovin Report, we estimate the maximum release rate +
of radiofodines to have occurred between 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br /> on Parch 28 and 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br /> on March 30,1979 at 2.26 x 10-5 Ci/sec measured as iodine-131, I
D.
Identify the postulated accidents in which the releases of radiat may be several orders of magnitude higher than the. Unit 2 acciden
- and identify the estimated quantities which could be released as as an estimate of the maximum release rate.
Response
The staff has identified three design basis accidents and provided the quantities and release rate for noble gas and iodine-131 in the follow-ing table. Accidents with more severe release rates are considered in the design basis accident evaluation.
TABLE I-1 31 floble Gas Standard Review I-1 31 Release Rate floble Gas Release Rate Plan Accident Guantity, Ci Ci/sec Quan'ity, ci Ci/sec loss-of-Coolant 6
(SRP 15.6.5) 600
.,003 2 x 10 4.3 Fuel Handling 4
(SRP 15.7.4) 39
.005 4 x 10 6.1 Gas Decay Tank Rupture (SRP 5
15.7.1)
Negligible Negligible 9 x 10 12.8 E.
Did the effluent monitoring system for Unit 1 display any off-scale readings durirg the Unit 2 accident:
If so, specify.
Res ponse No.
They were on-scale during the TMI-2 accident.
F.
Does Licensee now have in place at Unit 1 effluent monitoring systems which are capable of staying on-scale under conditions of a Unit 2 accident?
Response
Increased effluent monitor range capabilities will be provided for each
- of the gaseous effluent discharge paths prior to restart.
The licensee has submitted descriptions of a proposed monitoring system with the cap-ability to remain on-scale under conditions if a Unit 2 accident occurs for staff review.
G.
If not, will such systems be required as a precondition of restart if they will not, why not?
Response
Interim procedures for quantifying high level accidental radioactivity re-leases will be required as a condition of restart.
By January 1, 1981, permanent monitoring systems will be required.
Guiaance for the increased monitor range requirements is found in the October 30, 1979 letter to all licensees, subject " Discussion of Lessons Learned Short Term Requirements".
5 What is Staff's position regarding compliance of TMI-l with GDC 64?
Explain your answer.
Response
The staff considers that the modifications made to the gaseous effluent monitoriag instrumentation to the new increased range capabilities stated in our response to 5-1F and G are adequate, and can provide exasured release rates during postulated accidents in cc.1formance with 10 CFR 50, Appendix A, GDC 64.
5 What is Staff's position regarding when the high-range effluent monitoring system must be installed?
Response
The staff position is that the high-range effluent monitoring capability be installed prior to TMI-l restart in compliance with the recommendations of Section 2.1.8.b of NUREG-0578 as explained in the October 30, 1979 letter to all licensees. See the response to 5-1F and G.
Interrogatory 6-2 Does Staff plan to require design changes to TMI-1 to counter the five features listed in Contention #6?
If not, why not?
If so, specify.
Ep_s_ pons e The five features referenced in Contention #6 are discussed individually below.
Item a:
Design of the steam generators to operate with a relatively small liquid volume in the secondary side.
Because of the small water volume in the steam generators, as well as other factors, the staff found that B&W-designed reactors place more reliance on the reliability and performance characteristics of the auxiliary feedwater system.
This occurs because the B&W steam generators contain less total water inventory than those of the other reactor vendors, and therefore, need more rapid actuation of the auxiliary feedwater system for continued decay heat removal following the loss of main feedwater.
To increase the reliability and performance character-istics of the TMI-1 emergency feedwater system, so as to compensate for the low SG water volume, the staff will require a number of actions, including improved operating procedures, improved instrumentation, a safety-grade system for automatic initiation of emergency feedwater, and the capability to control emergency feedwater independently of the Integrated Control System, as described in the SER.
Item b:
Lack of direct initiation of reactor trip upon the occurrence of off-normal conditions in the feedwater system.
6-2-2
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The staff has required that an anticipatory reactor trip system be installed at TMI-1 which initiates a direct reactor trip upon loss of the main feedwater pumps or a _ turbine trip, as described in the SER.
Item c:
Reliance on an integrated control system to automatically regulate feedwater flow.
The staff has required that procedures be developed and a method be available for controlling emergency feedwater independently of the ICS. as described in the SER. -
Item d:
Actuation before reactor trip of a pilot-operated relief valve on the primary system pressurizer.
The Staff has required that the high pressure reactor trip set point be reduced to 2300 psig, and that the PORV opening set point be increased to 2450 psig, as described in the SER.
With these revised set points, the pilot-operated relief valve will not be actuated before reactor trip.
Item e:
Low steam generator elevation relative to the reactor vessel which provides a smaller driving head for natural circulation.
Operating guidelines have been developed which require raising the steam generator water level when forced circulation is lost.
Analysis has shown that core cooling can be maintained through normal and accident conditions by the natural circulation resulting from these water levels.
Additionally,naturaY circulation has been demonstrated at B&W plants similar to TMI-1, showing that sufficient flow rates are obtained to allow decay heat removal through the steam generators.
7.1 Does licensee now comply with 10 CFR S0 Appendix K requirements?
If not specify shortcomings.
The Licensee's ECCS evaluation model was reviewed by the Staff and found to be in compliance with Appendix K to 10 CFR 50 on January 8, 1976 and amended on September 5, 1978.
Recently, as part of the Bulletins and Orders Task Force Review, the Staff has identified potential shortcomings in the Licensee's small break ECCS evaluation model.
It is unknown at this time if any of these potential shortcomings will require model revisions.
The Task Force has recommended in tiUREG-0525 that the licensee be requested to submit the ECCS small break evaluation model with appropriate revisions, if any, by July 1,1980.
4 -
10 Has Staff evaluated the impact of Unit 2 decontamination procedures on.
activities at Unit l?
If not, why not?
If so, specify which activities ^
were evaluated and the outcomes of the evaluations.
Response
The staff is reviewing the TMI-l Restart Report submitted by Metropolitan Edison and preparing a Safety Evaluation Report for the restart. The staff evaluation must conclude that the TMI-2 decontamination and restoration operations will not affect safe operations at TMI-l before it can restart.
Complete separation of liquid, gaseous and solid radwaste treataent systems is required such that each unit will manage its own generated.;aste, monitor and control its own radioactive effluents, and report its own iiapact on the environs according to tne regulations and to meet the conditions set forth in criteria #4 of the Commission's Order and Notice of Hearing, dated August 9, 1979, for TMI-l restart.
10 Will total separation of Units 1 and 2 be required of the Licensee as a precondition of restart?
Response
The Commission's Order and Notice of Hearing dated August 9, 1979, criteria #4, required the complete separation and/or isolation of TMI-l and TMI-2 radioactive waste systems, specifically, the radioactive liquid transfer lines, ventilation systems and sampling lines. These separation modifications will be completed and disconnections made on shared services. The Fuel Handling area at TMI-l will be separated from the Fuel Handling Building by an additional wall.
Dampers will be in-stalled in the Fuel Handling Building Ventilation supply and exhaust
~
lines to prevent potential leakage paths between the Fuel Handling P
Building and the Auxiliary Building at TMI-1.
Therefore, the technical specifications for TMI-l have been written to require pre-restart tests and routine tests to be performed on the Auxiliary and Fuel Handling Building Ventilation System prior to any fuel handling.
In addition, the licensee has proposed to install an additional Fuel Handlina area floor ESF Ventilation System prior to the first refueling outage after restart as a long term method of assuring air flow control in the floor area.
The staff has reviewed the method and finds the plan acceptable.
Interrogatory 17-1 To what extent has Staff's accident analysis procedure included analysis of Class 9 accidents?
Response
The Staff's accident analysis procedure does not include analysis of so-called Class 9 accideni.s.
Further details will be provided in a supplement to the f1RC Staff position on the need to consider Class 9 events, to be filed with the Board on or before March 21, 1980.
m G
s
Interrogatory 17-4 To what extent have the following features been included in Staff's accident analysis for Unit 1:
A.
Multiple failures, especially multiple failures of engineered safety features.
B.
Deliberate acts of sabotage by insiders.
C.
Operator error in responding to accidents and transients.?
Response
A.
The staff's accident analysis for Unit 1, done before the TMI-2 accident, did not search for specific combinations of multiple failures that might be conceivable.
Instead, a more generalized approach was employed.
Electrical and fluid systems which are designated as engineered safety features must be designed to meet requirements set forth in 10 CFR 50 of the Commission's regulations and be qualified as necessary to perform their function in an accident environment.
These design requirements, in them-selves, provide a high degree of assurance that these systems will be capable of carrying out their mission when called upon.
Nevertheless, it is possible for failures to occur, and redundancy must be provided to assure that performance requirements can be met when needed. To assure such redundancy, the following steps are taken in the design review and the accident analyses:
First, an initiating event or an initiating failure is postulated.
Then possible consequential failures in fluid or electrical systems are identified and presumed to have occurred in such a fashion that _,
the associated systems or subsystems would be degraded or rendered incapable of providing a safety function.
Finally, the single failure criterion is applied to the remaining aggregate of systems which supply the requisite
17-4-2 safety function. This entails the arbitrary assumption of an additional failure at various points in that aggregate and a requirement that the desi be such that the safety function can still be carried out successfully.
In this sense, a number of failures are accounted for in analyzing the perform-ance capabilities of various engineered safety features.
Anplifying details are provided in the Staadard Review Plan, various Regulatory Guides, and in 10 CFR 50 and Appendices A and K thereto.
B.
Deliberate acts of sabotage by insiders are not assumed in connection with the analysis of accidents.
However, plant security does receive a high degree of emphasis. This includes measures to provide high assurance against successful acts of sabotage by insiders.
These include background investigations of employees to assure trustworthiness, and restricting access to critical plant areas.
General performance requirements are set forth in 10 CFR 73.55(a).
C.
Operator error in responding to accidents and transients was not considered in detail prior to the TMI-2 accident.
In some cases, operator errors of commission or omission were considered equivalent to single active failures in applying the single failure criterion.
In cases where such errors might occur, but time was available, reliance on corrective operator action was permitted.
However, if rapid actuation of an engineered safety feature were needed, actuation was required to be automatic.
-s In the period since the TMI-2 accident a number of actions have been initiated to account for multiple failures and operator errors and to improve the capability of plant personnel to diagnose and cope with off-normal situations.
17-4-3 A summary of such actions (completed, ongoing, or planned) will be included in a supplement to the NRC Staff position on the need to consider Class 9 events, to be filed with the Board on or before March 21, 1980.
Respectfully submitted, kd./&>f s
c Lucinda low Swartz Counsel for NRC Staff Dated at Bethesda, Maryland this 17th day of March, 1980 9
UtilTED STATES OF AMERICA fiUCLEAR REGULATORY COMMISSI0ff BEFORE THE ATOMIC SAFETY Af;D LICEf SIf;G BOARD In the Matter of
)
)
ME FROPOLITAff EDIS0f1 COMPAtlY, et _al.
)
Docket lio. 50-289 (Three Mile Island, Unit 1)
)
AFFIDAVIT OF GLEliff B. KELLY I, Glenn B. Kelly, being duly sworn, do depose and state:
l.
I am a reactor engineer in the Division of Systems Safety, Office of f!uclear Reactor Regulation of the United States f;uclear Regulatory Co: anis s ion.
I am responsible for reviewing accident analyses and ECC and RHR systems of assigned nuclear power plants.
2.
The answer to Steven C. Shelly's Interrogatory 2-2 was prepared by me.
I certify that the answer given is true and accurate to the best of my knowledge.
7 O
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Glenn t. Kelly '
Subscribed and sworn to before me this /U" day of ic5:
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f60tary PuO ic
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My Comission expires:
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s UNITED STATES OF AMERICA NUCLEAR REGULATORY C0:' MISSION BEFORE THE ATO'41C SAFETY AND LICENSING BOARD
~
In the Matter of
)
)
METROPOLITAN EDISON COMPANY, et al.
)
Docket No. 50-289
)
(Three Mile Island, Unit 1)
)
AFFIDAVIT OF BRIAN W. SHERON I, Brian W. Sheron, being duly sworn, do depose and state:
1.
I am a Principal Nuclear Engineer in the Division of Operating Reactors, Office of Nuclear Reactor Regulation of the United States Nuclear Regulatory Commission.
I am responsible for performing the safety review of assigned nuclear power plants, including Three Mile Island, Unit 1 Restart Program.
2.
The answer to Sholly Interrogatory 3-4 was prepared by me.
I certify that the answer given is true and accurate to the best of my knowledge.
- l. 44 ff/ks Fian W. Sheron Subscribed and sworn to before me this
/ day of March 1980.
lG2 nn lh. w(n
~-
w' a Notary Public
/0 My Commission expires:
I, / / 9[ "
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
METROPOLITAN EDISON COMPANY, et.al.
Docket No. 50-289 (Three Mile Island, Unit 1)
)
AFFIDAVIT OF BRIAN W. SHERON I, Brian W. Sheron, being duly sworn, do depose and state:
1.
I am a Principal Nuclear Engineer in the Division of Operating Reactors, Office of Nuclear Reactor Regulation of the United States Nuclear Regu-latory Commission.
I am responsible for performing the safety review of assigned nuclear power plants, including Three Mile Island, Unit 1 Restart Program.
2.
The answer to Steven C. Sholly Interrogatory 7-1 was prepared by me.
I certify that the answers given are true and accurate to the best of my knowledge.
f/ dfff Mt -
Brian W. Sher 6n Subscribed and sworn to before me this '/ < day of 9t% A-l?h cit.
L totaly Public y
J /, lV My Commission expires:
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UtlITED STATES OF AMERICA liUCLEAR REGULATORY C0: MISSION BEFORE THE ATOMIC SAFETY Af4D LICEf4SIf4G BOARD In the Matter of
)
)
METROPOLITAtt EDISOff COMPAt1Y, e._t. _al_.
)
Docket f;o. 50-289 (Three Mile Island, Unit 1)
)
AFFIDAVIT OF RICHARD E. IRELAr?D I, Richard E. Ireland, being duly sworn, do depose and state:
1.
I am a Technical Advisor in the I;uclear Regulatory Commission Staff's Division of Systems Safety.
In that assignemt I perform a variety of tasks related to resolution of safety problems on power reactors of all types.
I am temporarily' assigned to the Division of Operating Reactors to assist in establishing a program for review and feedback of safety-related information from operating experience at nuclear power plants.
2.
The answers to Sholly Interrogatories 17-1 and 17-4 were prepared by me.
I certify that the answers given are true and accurate to the best of my knowledge.
t l
Richard E. Ireland 1
Subscribed and sworn to before me this / 7 day of e
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I,otary Public
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- y Commission expires:
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