ML19340B120
| ML19340B120 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 10/07/1980 |
| From: | Trimble D ARKANSAS POWER & LIGHT CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0660, RTR-NUREG-660 1-100-02, 1-100-2, 2-100-05, NUDOCS 8010210454 | |
| Download: ML19340B120 (8) | |
Text
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ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551 UTTLE ROCK, ARKANSAS 72203 (5011371-4000 October 7, 1980 l'-100-02 2-100-05 Mr. Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Subject:
Arkansas Nuclear One - Units 1 and 2 Docket Nos. 50-313 and 50-368 License Nos. DPR-51 and NPF-6 Clarification of TNI Action Plan Requirements (File:
1510.3, 2-1510.3)
Gentlemen:
In our attempts to meet the NRC requirements which have developed from the TMI-2 incident, there has been a need for an "all encompassing" document which would list all the TMI-2 incident related requirements and implementation dates.
The subject document is a step in the right direction toward achievement of this goal.
However, some concern has arisen within AP&L over the new requirements which were imposed by the document.
While relief was given on many of the due dates, the scope of many of the requirements was significantly increased.
Specific examples of the increased scope and the difficulties involved in meeting them are given in an attachment to this letter. Some general comments as to the increased scope of the requirements are given below.
1)
The use of Regulatory Guide (RG) 1.97 (Revision 2) as the basis of the design requirements for the instrumentation is inappropriate.
RG 1.97 (Rev. 2) is still in draft form and has not been approved by the Commission and, therefore, does not provide an appropriate basis for a commitment of millions of dollars and thousands of man-hours of effort by the utilities.
Changes to the RG before approval could result in the design, qualification and installation of new equipment that is either not adequate or in excess of the final re-quirements of the RG.
2)
There are many instances where the staff has shown a major interest in the human factor element of control room design (e.g., NUREG/
CR-1580, NUREG-0660, NUREG/CR-1270).
However, many requirements for additional instrumentation (e.g., containment pressure, water 4@
8010210 MEMBER Mif ' OLE SOUTH UTIUTIES SYSTEM
level, hydrogen monitors) and new systems (e.g., in-containment radiation monitor, high-range effluent radiation moniter, Safety Parameter Display System (SPDS), TSC instrumentation, Safety Grade Emergency Feedwater Initiation and Control System, addi-tional instrumentation for detection of Inadequate Core Cooling, etc.) have been made with little consideration for human factor concerns. While many of the individual requirements have the potential to improve plant safety, when considered collectively the impact on the control room and the operator is enormous.
Putting this much additional. instrumentation in the control room has the potential for reducing plant safety by cluttering the control room with instrumentation not required for normal oper-ation and most transients.
3)
The requirement for Environmental Qualification of Equipment is not a TMI-2 incident related requirement and is, therefore, in-appropriate in this document.
The schedules and objectives for this requirement are not compatible with the TMI-2 incident re-lated requirements.
The Environmental Qualification of Equipment requirement should, therefore, be incorporated into the efforts of the NRC Environmental Qualification Branch.
4)
We do not feel that proposed Technical Specifications (TS) should be submitted along with the documentation for those items which require pre-implementation review.
Current TS require review by the Plant Safety Committee and the Safety Review Committee prior to submittal to NRC.
NRC rejection of;our design proposals could alter the proposed TS and thus cause an iteration through our re-view and NRC review.
The systems designs must be finalized before appropriate TS can be written.
5)
Responses to several items weredue by October 1,1980, as given in the draft clarification letter.
It was our understanding from the regional meeting held on September 24, 1980, that a blanket pardon was given for those due dates.
Therefore, no responses will be made on these items until the final approved version of the clarification letter is issued for implementation.
We hope the comments made in this letter and the attachment to it will receive appropriate consideration in revising the clarification letter.
Please contact us if we may further clarify our comments or positions.
Very truly yours, h5 David C. Trimbl Manager, Licensing DCT:g:s1 Attachment f
y
I.A.1.1 SHIFT TECHNICAL ADVISOR Our current STA program is geared to the NRC requirements, not the recommendations of the INP0 document.
The INPO recommendations go significantly beyond the NRC requirements.
For example,100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of simulator training is recommended by INPO, whereas only 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />
.is required by NRC guidelines.
Time on the simulator has already been scheduled by AP&L and other utilities for the next two years.
Obtaining additional time would be difficult, if not impossible.
I,A.3.1 REVISE SCOPE AND CRITERIA FOR LICENSING EXAMINATIONS No clarification is given as to what the simulator exam will include.
Will it include only a startup, casualty, failures? Will oral ques-tions be asked during the exam? Who will be in charge of conducting the exam and arranging for time on the simulator? These questions need to be answered and guidelines set up before implementation of the program, i
I.C.1 PROCEDURES FOR TRANSIENTS AND ACCIDENTS We are not in receipt of letters dated September 27, October 10, or November 9, 1979, as referenced in the clarification. The October 30, 1979, letter only acknowledges this issue and the' September 13, 1979, letter merely implements NUREG-0578.
NUREG-0578 contained a half page explanation of this item that indicated computer calculations were only needed where system response to operator actions were unclear or where quantitative information important for input to operator instructions could only be obtained in that manner.
Natural phenomena were not men-tioned and no mention of submittal to NRC of any information in this area was made.
In addition, consideration of multiple failures was given the status of a future possibility.
In short, the " clarification" provided in I.C.1 is a new requirement.
In fact, claims that "NUREG-0578 concluded that the single failure criteria was not considered appropriate for guideline development and called for the consideration of multiple failures" are not correct.
As the NRC staff is well aware, tremendous efforts have been expended on.the original NUREG-0578 requirement.
Implementation of the "clari-fication" in I.C.1 would require a new effort approximating or exceed-ing the effort already expended, which has consumed in excess of one year and promises to continue for another six months to a year or more.
Aside from the appropriateness -of-the new requirements themselves, experience with the original requirement 3 makes it obvious that the implementation schedule for the new requirements is unrealistic.
As to the appropriateness of the new requirements, we feel that inclu-sion of ATWS concerns before the ATWS issue is settled would only con-fuse both ATWS and the new requirements.
In addition, we feel that the movement of the staff into the area of procedure' review is both inadyis-able and a waste of staff resources since the procedures are of necessity very plant specific and subject to frequent changes.
In fact, we do not
believe that the adequacy of a generic guideline can be properly deter-mined by the staff without the guideline being so general as to be worth-less because of the substantial differences in the plants involved.
The more practical and useful approach is that taken in the original require-ments which called for no submittal, only action.
In short, I.C.1 is a new requirement that requires much more time and ef!. rt than apparently perceived by the promulgator and moves into areas thas have a significant potential for being counter productive.
II.B.1 REACTOR COOLANT SYSTEM VENTS As was indicated by AP&L staff members at the regional meeting on Septem-ber 24, 1980, no specific qualification requirements were planned on being imposed here.
There are no standards mentioned to which the equipment must be qualified.
Direction must be provided to allow an adequate response to this new requirement.
4 II.B.2 PLANT SHIELDING As stated in our letter, we do not believe this to be the appropriate place for environmental equipment qualification requirements related to existing equipment. These NRC concerns should be addressed through the efforts of the NRC Environmental Qualification Branch.
We also take exception to including those components on systems required to achieve cold shutdown in the qualification requirements.
Both ANO-1 and ANO-2 were designed under the criteria to achieve hot shutdown. Reg-ulations in force at the time of receipt of the Operating Licenses for both units specified hot shutdown as safe shutdown.
We contend, therefore, i
that there is no basis for the " requirement" to qualify equipment necessary to achieve cold shutdown.
Per the regional meeting on September 24, 1980, the January 1,1981, date given in Item 4 should be January 1,1982.
II.B.3 POST ACCICENT SAMPLING CAPABILITY Item 3.k requires backup sampling through grab samples if in-line monitor-ing is used.
Per the regional meeting, it is our understanding that tne analysis capability can be performed offsite provided ti,o time constraints are met.
This should be stated in the requirement.
II.E.1.2 EFW AUTO INITIATION AND FLOW INDICATION The ANO-2 design is in full compliance with the current _ requirement, while the ANO-1 system has been upgraded to meet the Category A NOREG-0578 re-quirements (i.e. control grade initiation and control separated from the ICS).
Safety grade flow indication will be fully installed during the up-coming refueling outage. AP&L, Florida Power Corp., and Sacramento Municipal
. l l
Utility District presented a comprehensive plan and schedule for EFW j
system upgrade to the NRC staff on September 4,1980.
This presen-l tation included our plans for the design, equipment procurement, and l
implementation of an EFW system upgrade for each plant.
Even though l
the NRC imposed due dates will not be met, we feel that the delay in l
implementation will result in a much more reliable EFW system.
II.E.4.2 CONTAINMENT ISOLATION DEPENDABILITY A decision to require the use of radiation as a third diverse contain-ment isolation signal will require at least two years for implementa-tion.
At the present time, appropriate detectors (low to medium range) which meet the required qualification standards do not exist.
Because l
the system would be an Engineered Safeguard system which required test-l ability, a minimum of three separate channels wculd be required. The l
setpoint for each channel would be dependent on the radiation level at the monitor location, thus requiring an extensive analysis of contain-j ment radiation levels.
This will probably result in a different set-l point for each monitor, further complicating the design, installation l
and operation of the system.
Our responses to IE Bulletins 79-05, 79-06 and the subsequent amendments stated our definition of essential and nonessential systems.
Should the NRC staff not agree with our d6finition, their concerns should be ad-dressed directly to AP&L.
It should be noted that any additional equip-ment would have to be Class IE, thus resulting in lead time of at least one year.
The new requirement for lowering the containment pressure setpoint to "the minimum compatible with normal operating conditions" would require i
an analysis of the containment pressure during normal operation. Because l
this analysis was performed during plant licensing to determine the initial setpoint, AP&L contends that we are in compliance with the current requirements.
l II.F.1 (ATTACHMENT 1)
NGBLE GAS EFFLUENT MONITOR l
Meeting the requirement as stated would necessitate the use of Class 1E l
equipment, which is not available.
Because there can be no test data for nonexistent equipment, further clarification is needed as to how a l
verification of the seismic criteria can be performed.
1 II.F.1 (ATTACHMENT 2)
SAMPLING AND ANALYSIS OF PLANT EFFLUENTS The requirement for a 30 minute sampling time for gaseous radioiodine and parti,culates at' concentrations of 10 uCi/cc would result in suf-ficient personnel exposure levels so as to be contradictory to the
.ALARA program.
It is AP&L's position that personnel exposure, and therefore sampling time,_should be kept to a minimum.
l The requirement to maintain the humidity at less than 80% will result in inaccurate measurements of radiciodines and particulates.
Efforts to reduce the humidity will also result in a reduction of the radio-nuclides, the extent of which cannot be ace:urately determined.
The net result will be an underestimation of the radionuclides to the en-v.ironment, thus potentially affecting the margin of safety provided l
to the public.
II.F.1 (ATTACHMENT 3)
CONTAINMENT HIGH RANGE RADIATION MONITOR Consideration should be given to the availability of monitors capable of covering the required range (1 R/hr to 10 R/hr).
Preliminary in-l formation has revealed only one vendor with a monitor which can cover l-this range.
The accuracy requirement of + 20% for photons between 0.1 MeV and 3 MeV may not be. achievable, particularly at the lower end of the range.
We are presently attempting to obtain response accuracy information from two different vendors.
Presently, only two venders are attempting to qualify their containment j
high range radiation monitors. Neither vendor has any provisions for l
using a calibrated radiation source for the lower end of the range. All l
calibration is currently accomplished by electronic signal substitution.
The "Special Environmental Qualifications", which are presented as new requirements, place very difficult restraints on the two vendors which are currently attempting to qualify the monitor.
Failure by these vendors to meet your qualification requirements would result in no utility being able to provide a monitor which meets your requirements.
Consideration should be given to the possibility that these requirements are beyond the existing " state of the art" for this equipment.
l II.F.1 (ATTACHMENT 4)
CONTAINMENT PRESSURE MONITOR j
The currently installed containment pressure instrumentation is used as l
inputs to the Engineered Safeguards actu: tion system.
Any changes in l
monitor accuracy which result from expanding the range from 70 psia to l
180 psia would require setpoint changes and safety analysis reviews The present instrumentation has two indicators, one per channel, v a
one channel recarded.
Accuracy requirements would necessitate the use of two separate row-fed
)
pressure instrumentation channW s.
Regulatory Guide 1.97 requires that the instrumentation be located in the Control Room, with redundant chan-nels.
Also, one channel must be recorded.
As a result of imposing this requirement, Containment Pressure would be indicated ~on four (4) indicators and two (2) recorders.
This further clutters the Control Room panels and could result in operator confusion.
Again, Human Factors concerns were not factored into this requirement.
~5-II.F.1 (ATTACHMENT 5)
CONTAINMENT WATER LEVEL MONITOR NUREG-0578 required that the wide range water level instrumentation be designed and installed in accordance with Regulatory Guide 1.97.
The narrow range water level instrumentation was to meet the requirements of Regulatory Guide 1.89.
The clarifications stated that both narrow and wide range instrumentation must meet Regulatory Guide 1.97.
This is a significant change in requirements for the narrow range instru-mentation, as new equipment would have to be procured.
This equipment procurement will require approximately six months to one year, however no extension of the due date was given.
The benefits of this require-ment should be assessed to determine its value in improving safety 4
versus its impact on Human Factors concerns in the control room.
The clarifications have increased the documentation significantly over the NUREG-0578 requirements.
The new requirements are: (1) specify the elevations of safety-related equipment relative to the maximum water level in the containment and (2) specify the minimum water level in the containment which satisfies pump NPSH requirements.
Item (1) will re-quire a significant engineering effort to determine the location and relationships of all instrumentation to the maximum water level.
This effort should be incorporated into the IE Bulletin 79-01B effort because of its similarities to that effort.
Item (2) will require an engineer-ing evaluation of pump specifications.
This effort can be provided on the current schedule.
II.F.1 (ATTACHMENT 6)
CONTAINMENT HYDROGEN MONITORS NUREG-0578 required that the Hydrogen Monitors be capable of measuring hydrogen to a concentration of 10% and be qualified to Regulatory. Guide 1.97 (Revision 1) which requires equipment to be qualified to IEEE-323-1971 and IEEE-344-1971.
The clarification essentially has five (5) new requirements above the NUREG-0578 requirements.
Clarification (1) requires that the instru-mentation be designed and qualified to Regulatory Guide 1.97 (Revision
- 2) for Category 1 instruments.
The equipment purchased Tor NUREG-0578 requirements meets Regulatory Guide 1.97 (Revision 1).
This is a sig-nificant upgrade in the equipment qualification requirements, since Revision 2 requires that equipment be qualified to IEEE-323-1974 and IEEE-344-1975.
To meet Regulatory Guide 1.97 (Revision 2), new eeuip-ment would be required. The lead time for Hydrogen Monitoring Instru-mentation qualified to IEEE-323-1974 is approximately one (1) year.
The cost of each monitor is approximately $60,000, plus the cost of the installation, which will bring the total cost for implementation of this requirement to approximately $200,000 per Hydrogen Monitor Channel.
Clarification (2) is a relief in requirements'and can now be accomplished with equipment that is currently designed and available for purchase.
Clarification (3) requires an accuracy of 1 1% of the monitored range.
The qualified Hydrogen Monitors typically available have an accuracy of 1 2% of applicable range, repeatability of i 1% of applicable range, re-producibility 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s: 11.5% of range, zero and span drift: 1 2% in
one (1) week and speed of response:
90% of reading in less than 60 seconds.
Clarification (4) is a very significant change from the NUREG-0578 requirements.- The analysis to determine the appropriate number of sample points and the design, procurement and construction necessary to implement the sampling system is a very significant effort.
The analysis could take as much as six (6) months.
The de-sign would involve installation of seismic piping and Class 1E valves.
The design could be a significant effort, depending on the number of sample points and the sample locations.
Class 1E valves normally re-quire one (1) to one-and one-half (1 ) years to procure.
The documen-tation requirements as given in the clarifications will be difficult, if not impossible, to provide for the hydrogen monitor installed in the plant at this time.
Specifically, items b, c and d may not be available for the Hydrogen monitors procured for our ANO-1 and ANO-2 plants since this equipment was purchased in the late 1960's and early 1970's.
The best available information on limiting operating conditions, performance sensitivities to environmental conditions and testing for post-LOCA conditions will be provided.
II.F.2 INSTRUMENTAi10N FOR DETECTION OF INADE0VATE CORE COOLING AP&L has previously provided to the NRC staff the results of our analy-sis which determined that no additional instrumentation is needed.
Also, procedures have been provided to the staff which address Detec-tion of Inadequate Core Cooling, using the existing instrumentation.
AP&L has recently received a letter on this subject from Darrell G.
Eisenhut dated September 24, 1980.
Our concerns and comments will be provided in response to the Eisenhut letter of September 24, 1980.
II.K.3.7 EVALUATION OF PORV OPENING PROBABILITY This is a new requirement which will require extensive analyses.
The requirement is essentially to prove "that the PORV will open in less than five percent of all anticipated overpressure transients...."
However, no justification whatsoever is given for the five percent number.
Furthermore, it is apparently the desire of the NRC staff to require us to upgrade the PORV (II.K.3.1) and at the same time assure the staff that the PORV will hardly ever be used.
These requirements are contradictory.
The analysis required by II.K.3.2 (and any modifi-cations which result from the analysis) should answer NRC staff concerns about the PORV and negate the need for an evaluation of the PORV opening probability.
i l