ML19312E273
| ML19312E273 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 06/02/1980 |
| From: | Sohinki S NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Doherty J DOHERTY, J.F. |
| References | |
| NUDOCS 8006040115 | |
| Download: ML19312E273 (18) | |
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06/02/80 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
HOUSTON LIGHTING & POWER COMPANY
)
Docket No. 50-466
)
(Allens Creek Nuclear Generating
)
Station, Unit 1)
)
NRC STAFF'S SECOND PARTIAL RESPONSE TO JOHN F. DOHERTY'S TWELFTH SET OF INTERROGATORIES The NRC Staff responds, in part, as follows to the twelfth set of interrogatories propounded by John F. Doherty in the captioned proceeding.1/ By agreement with fir. Doherty, the remaining responses will be filed as soon as the necessary Staff reviewers complete current review assignments.
12-10-01. According to Nuclear Safety 20(2),175,1977, "... the starting systens rer.;:n the najor problen area" and
" reliability of on-site diesel generators was less than anticipated."
a.
What is planned to improve starting reliability in licensees?
b.
Has applicant's planned any aspects that in staff's opinion will improve reliability in:
1.
Starting 2.
Fuel system (cut down on cir,qging, etc.)
3.
Lube oil system (over-heatirig, bearing burn-out, etc.)
4.
Cooling 5.
Soeed control (governor reliability) 6.
Speed control circuitry 3 The Staff's first partial response to Mr. Doherty's twelfth set of interrogatories was filed on May 20, 1980.
8006040l\\
Response
a.
NUREG-CR-0660, " Enhancement of Onsite Emergency Diesel Generator Reliability" made specific recommendations for increasing the reliability of nuclear power plant emergency diesel generators. These recommendations are based on comprehensive study of diesel generator operating experience at nuclear power plants and on consultations with major diesel generator manufacturers. These recommendations address the following areas:
1.
Moisture in the Air Start System 2.
Dust and dirt in D/G room 3.
Turbocharger gear drive problem 4.
Personnel training 5.
Automatic prelube 6.
Testing, test loading and preventing maintenance 7.
Improve identification of root cause of failures 8.
D/G ventilation and combustion air systems 9.
Fuel storage and handling
- 10. High temperature insulation for generator 11.
Engine cooling water temperature control
- 12. Concrete dust control
- 13. Vibration of iratruments and controls All operating plants and plants in the licensing process are being required to review their designs for conformance to these recommendations, and to make design and procedural changes in this regard as deemed necessary by the staff to enhance the overall reliability of the diesel generator units.
b.
The Applicant for Allens Creek will be required to implement the reconinen-dations in its design as cited in item (a) above. Therefore, the overall reliability of the diesel generators will be improved.
12-10-02.
Does the NRC do a quality assurance inspectioq on General Motors Generator units like applicant plans ta use? If so, does it include (a) governor, (b) governar circuitry.
Response
The NRC's Office of Inspection and Enforcement conducts periodic scheduled and unannounced field inspections of the Applicant's quality assurance program implementation as well as those of its contractors and suppliers.
So far, General Motors, supplier of diesel generator (HPCS) for Allens Creek is not included in the list of vendor's tnat NRC inspects.
The Applicant's quality assurance program which is submitted for our review, however, includes quality assurance requirements for diesel generator auxiliaries including diesel generator governor. The governor circuitry is verified through pre-operational and periodic testing of the diesel generator units. These tests are witnessed by our resident inspector.
12-10-03.
In NUREG-0572, p. D-44, ACRS states, " correctable problems l
(e.g., associated with maintenance) can be identified from j
the LER data both on a generic and specific plant basis."
\\
(a) What problems can be identified with G. M. diesel generators from LER's?
Response -
There are 27 LER's on General Motors diesel generator events from 1969 to the present. The following are the most significant problems identified with General Motor diesel generators from the review of these LER's:
1.
Moisture in the starting systems.
2.
Dust and dirt in diesel generator compartments.
3.
Air in diesel fuel lines.
4.
Fuel leak caused by failure of tubing from fuel line tc ftiel pressure gage.
5.
Turbochargers gear drive problems.
6.
Personnel error during periodic testing.
It should be noted that except for item 5 the problems identified above on G. M. diesel generators are not significantly different than those on diesel generator supplied by other manufacturers. The implementation of the recommen-dations of NUREG-0660 as described in the response to Interrogatory 12-10-01 will effect the necessary corrective measures in the above cited problem areas.
12-10-04. When will applicant be required to determine the manufacturer of its diesel generators for balance of plant safety systems?
(that is for other than HPCS).
Response
There is no safety criterion that requires the Applicant to determine the manufacturer of its diesel generator units at the constructicn permit stage.
However, by the operating license review stage, the diesel generator manu-facturer is already determined by the Applicant since the Applicant is required
to submit the results of the diesel generator reliability qualification testing for NRC review. The qualification testing requirements are provided in IEEE std 387-1977 " Standard Criteria for Diesel Generator Units Applied as Standby Power Supplies for Nuclear Power Generating Stations," as augmented by Regulatory Guide 1.9 Revision 2.
12-10-05.
In NUREG-0572, pg. D-44, ACRS points out a disparity between test conditions and " cold start" for diesel generators, without elaboration. What disparaties are they referring to?
Response
The disparities ACRS pointed out in NUREG-0572, p. D-44 between test conditions and " cold start" for diesel generators are as follows:
1.
During diesel generator testing, the lube oil system is normally prestarted until satisfactory lube oil pressure is established in the engine main lube distribution header. This is done to reduce bearing wear, since long periods on standby mode have a tendency to drain the engine lube oil piping system. This prelube is not done during emergency starting.
2.
During monthly testing, the diesel generators are manually started and loaded.
During 18 months periodic testing however, the essential loads are energized through the load sequencer to test the actual sequencing of loads. During accident conditions, the safety loads are sequentially loaded on the safety buses.
12-10-06. What steps must applicant take by requirements of the Commission if one diesel generator to HPCS is out of service?
N'
. Response The standird Technical Specifications for BWR's require that if the HPCS diesel generator is inoperable, the operation of the plant could continue if the availability of the remaining power sources is verified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.
If these conditions for continued power operation are met and the HPCS diesel generator is restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unrestricted operation may resume.
If the conditions for continued power operation are met and the HPCS diesel generator is not restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the HPCS system will be declared inoperable.
For HPCS System inoperable, the operation of ECCS divisions 1 and 2, the ADS and the RCIC System will be verified.
If these conditions are met, power operation may continue provided the inoperable HPCS system is restored within 11 days.
If the HPCS system is not restored within 11 days, the unit would be brought to a hot shutdown state within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
12-10-07. What steps must applicant take by requirements of the commission if one diesel generator to the balance of plant safety systems is out of service?
Response
See response to interrogatory 12-10-08.
12-10-08. How long will applicant's reactor be permitted to operate when there is but a single incoming power source line?
Response
The above degradation level means that the available A.C. power sources (offsite power circuits or diesel generators) are one less than the limiting n.
conditions for operation (in accordance with Regulatory Guide 1.93, the limiting conditions for operation with respect to available electric power sources is an electric power system that satisfies General Design Criterion 17
" Electric Power Systems," by including the following electric power sources:
(1) two physically independent circuits from the offsite transmission network, (2) redundant onsite AC power supplies, and (3) redundant onsite DC power supplies). Thus, in accordance with Technical Specification requirements operation of the plant could continue if the availability of the remaining sources is ve-ified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.
If these conditions for continued power operation are met and the affected source is restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unrestricted operation may resume.
If the conditions for continued power operation are met and the source is not restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the unit would be brought to a hot shutdown st-te within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to a cold shutdown state within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
12-10-09. Has NRC requested 1 review of IEEE std. 387-1977 in view of dissatisfaction sith reliability of starting in nuclear plant generators?
Response
i The NRC has not requested IEEE committee revision of IEEE Std. 387-i 1977. However, as stated in our response to interrogatory 12-10-01, we are in the process of implementing the recommendations contained in NUREG-CR-0660, " Enhancement of Onsite Emergency Diesel Generator Reliability," in l
l
I all light. water operating plants and all plants in the licensing process.
We believe that implementing these recommendations is the most direct and effective way to upgrade the reliability of the diesel generator units.
12-10-10. What steps are licensees required to take to insure that spurious actuation of generator protective trip systems does not prevent the diesel generator unit from performing its function?
(Include anything from 1977 on in your answers, and it may assist to see page 1.9-2 of Reg. Guide 1.9 Rev.1).
Response
In accordance with position C(8) of the Regulatory Guide 1.9 Revision 2, we require that with the exception of the engine overspeed trip and the generator differential trip, all diesel-generator protective trips should be either (1) implemented with two or more independent measurements for each trip paramater with coincident logic provisions for trip actuation or (2) automatically bypassed during accident conditions. This will insure that spurious actuation of generator protective trips does not prevent the diesel generator unit from performing its function.
=
12-17-04. Has staff any data from a source other than G.E. for the statement, "It is our understanding that the newly designed actuator has by tests demonstrated improved reliability due (to) the elimination of the bellows and its renaced sensitivity to pilot valve laakage."?
(Page 3 of memo B. K. Grimes (NRC) to all BWR Licensees on 4/5/79, #7905090171.
If so, what sources:
Response
The Staff's review has relied on data from G.E.
The Staff also recognizes that during development of new designs, data may be generated for use in evaluations by the manufacturer and G.E without the data being submitted to NRC for its review. You should note that the letter referenced in the inter-rogatory refers to target rock safety-relief valves which are pilot actuated valves. As noted in Section 5.2.2 of the Safety Evaluation Report the valves in the Allens Creek facility will be balanced type, spring loaded safety valves provided with an auxiliary power actuated device which allows opening of the valve even when pressure is less than the safety-set pressure of the valve.
12-17-05.
Is the suppression pool factor of safety of 1.6 against yielding calculated strong enough to maintain integrity against a.
Turbine trip b.
Turbine trip with condensor by-pass failure c.
Transient due to heater failure d.
MSIV closure e.
Rod drop accident f.
Turbine trip without bypass (low power) g.
Loss of Condensor vacuum h.
Generator load rejection without by-pass 1.
Loss of auxilliary power and if "yes" indicate the margin between the transient conditions and the factor of safety if possible.
(Note:
See your reply to this Intervenor's Interog 6-2.)
~
Response
The transients and accidents that you have tabulated (12-17-05 a through 1) are all analyzed in Chapter 15 of the PSAR. As noted on page 15.1-4 of the PSAR, events considered in Chapter 15 of the PSAR include transients, design basis accidents, and intermediate events, and that containment stresses need not be less than those allowed for accidents by applicable industry codes except when containment is required. Thus, the Staff does not require a particular factor of safety for all of the events listed in Chapter 15 and hence, does not review the factor of safety for every event.
The design criteria for the suppression pool are described in Sections 3.8.2 (containment) and 3.8.3 (drywell).
For example, and for the steel containment shell, applicable codes, standards and specifications are listed in Section 3.8.2.2, loads and load combinations are given in Section 3.8.2.3, and stress limits are discussed in Section 3.8.2.6 and tabulated in Table 3.8.1.
Since discharge of all SRV's is a design condition, assurance of acceptable design margins is assumed for events of lesser suppression pool loads even though specific margins have not been identified and reviewed.
As stated in Section 3.8.1 of the Safety Evaluation Report, Supplement No. 2, i
we had concluded that confonnance with criteria to be used in the design of the containment constitutes an acceptable basis for satisfying in part the requirements of Criteria 2, 4,16 and 50 of the General Design Criteria.
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12-43-01: What is General Electric Alternative proposal 6.6 to Reg. Guide 1.54, as in a letter from S. A. Varga (NRC) to G. G. Sherwood (GE) of 2/1/77.
Response
No General Electric Alternatives to Regulatory Guide 1.54 are applicable to Allens Creek. On page C1.54-1 of the PSAR, Am. 46 7/14/78, the Applicant states that it will comply with the regulatory position of the guide. The Applicant did not request any review and approval of alternatives by the Staff.
12-44-03. What is the general NRC evaluation of the Internal Friction Nondestructive Evaluation Technique for Detecting Incipient Cracking in Pipes? Is the method considered superior to 4
accoustic emission technique?
Is there considered any prospects for using the method in nuclear applications where thc heavy radioactivity of much piping makes dealing with such pipes difficult? (You may wish to see 7909270409 for more on this)
Response
Research is being conducted by Daedalean Associates, Inc., on the application of the internal friction nondestructive evaluation technique for detecting incipient cracking of bypass lines and pipes in boiling water reactors piping systems. This research is in process and NRC evaluation on its usefulness and applicability has not been made. Whether the method is superit.r to the accoustic emission technique is not known at this time.
- l
12-44-04.
Is the " Spot Heating technique" one used only for piping before it is installed in a nuclear plant, or can the technique be used as a post-installation surveillance technique? (You may wish to see 7904173076)
Response
The spot heating technique has not been evaluated by the NRC Staff.
12-44-05. Will all ACNGS Coolant Pressure Boundary Piping be required to meet the guidelines of Part II of NUREG-0313, Rev.1, October 1979? If not, please indicate if it is too early to answer or there are some parts of the ACNGS plan which will require exception.
Response
The recommendations of NUREG-0313, Rev.1, will be implemented at the Allens Creek facility.
On page 1-29 of Appendix C to Supplement No. 2 to the Safety Evaluation Report we stated, "Because the Allens Creek facility is already in accordance with NUREG-0313, and because only nominal changes in requirements, e.g., expansion of augmented inspection programs to other areas, are expected to result from Task A-42, we conclude there is reasonable assurance that Allens Creek Unit 1 may be constructed and operated without endangering the health and safety of the public." Revision 1 to NUREG-0313 describes those expansions of augmented inspection programs which will be incorporated in the Allens Creek facility.
12-20-05. Does staff believe increasing FGR rates above 30,000 mwd / tonne might be observed as fuel temperatures decrease with burn-up?
J
Response
Under most conditions, increasing fission gas release rates would be observed above 30,000 mwd /MtM even though fuel temperatures may decrease with burnup. Temperature is the more dominant of the two major release Jependencies (temperature and burnup). At lower burnups, an increase in fission gas release rate would be expected only for increasing, constant, or very gradual decreases in fuel temperature. At higher burnups, particularly those above 40,000 mwd /'it.'1 the burnup dependence of fission gas release becomes stronger.
For these high burnup conditions, only very large reductions in fuel temperature would result in decreasing fission gas release rates.
12-20-06. By how much does fuel temperature decrease with burn-up between (a) 0-10,000 mwd / tonne U.; (b) 10-20,000 mwd / tonne U.
(c)20-30,000 mwd / tonne U.; (d) 30-45,000 mwd / tonne U.?
Response
Because fuel temperatures depend on a large number of specific conditions (e.g., fuel design, operating conditions, reshuffling scheme), it is difficult to quantify fuel temperature decrease with burnup. However, a general indication of how fuel temperatures vary with burnup can be found in Figure 1.
This figure shows GAPCON-2 predictions of fuel average temperature as a function of burnup for several different constant power levels. Power maneuvering, fuel depletion and other effects were not considered in these calculations.
It can be seen from the figure that the simplified constant power burnup history does not lead to monotonically decreasing fuel temperatures.
~
12-20-10. What is the amount of reduction of FGR calculated for 30,000 mwd /t and 45,000 mwd /t exposure when the operating temperature is reduced from 4,500'F to 3,300 F as occurred when ACNGS shifted from 7x7 fuel to 8x8 fuel? (That is, use the two burn-up points as fixed points; FGR is the dependant variable).
Response
The reduction in fuel temperatures that results from the shift from 7x7 l
to 8x8 fuel with2 a corresponding reduction in fuel rod power produces a corresponding reduction in FGR of about 15% FGR at high burnups (26% to 13% at 30,000 mwd /t and 46% to 30% at 45,000 mwd /t according to sample calculations we have on hand).
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMTSSIO'i BEFORE THE ATOMIC SAFETY AflD LICENSI!!G BOARD
~
In the Natter of HOUSTON LIGHTIllG & POWER COMPAflY Docket No. 50-466 (Allens Creek Nuclear Generating Station, Unit 1)
AFFIDAVIT OF CALVIN W. MOON I hereby depose and say under oath that the foregoing NRC Staff responses to interrogatories propounded by John F. Doherty were prepared by me or under my supervision.
I certify that the answers given are true and correct to the best of my knowledge, information and belief.
Calvin W. Moon Subscribed and sworn to before me this 2nd day of
- June, 1980.
Notary uFl'ic
/
& Ccamission expires:
July 1, 1982
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of HOUSTON LIGHTING & POWER COMPANY Docket No. 50-466 (Allens Creek Nuclear Generating
),
Station, Unit 1)
)
CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF'S SECOND PARTIAL RESPONSE T0 JOHN F. DOHERTY'S TWELFTH SET OF INTERR0GATORIES" and " AFFIDAVIT OF CALVIN W. MOON" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or, as indicated by an asterisk, through deposit in the Nuclear Regulatory Commission's internal nail system, this 2nd' day of June,1980:
Sheldon J. Wolfe, Esq., Chairman
- Richard Lowerre, Esq.
Atomic Safety and Licensing Board Panel Asst. Attorney General for the U.S. Nuclear Regulatory Commission State of Texas Washington, DC 20555 P.O. Box 12548 Capitol Station Dr. E. Leonard Cheatum Austin, Texas 78711 Route 3 Box 350A Watkinsville, Georgia 30677 Hon. Jerry Sliva, Mayor City of Wallis, Tex'as 77485 Mr. Gustave A. Linenberger
- Atomic Safety and Licensing Board Panel Hon. John R. Mikeska U.S. Nuclear Regulatory Commission Austin County Judge Washington, DC 20655 P.O. Box 310 Bellville, Texas 77418 R. Gordon Gooch, Esq.
Baker & Botts Mr. John F. Doherty 1701 Pennsylvania Avenue, N.W.
4327 Alconbury Street Washington, DC 20006 Houston, Texas 77021 J. Gregory Copeland, Esq.
Mr. and Mrs. Robert S. Framson Baker & Botts 4822 Waynesboro Drive One Shell Plaza Houston, Texas 77035 Houston, Texas 77002
~
Mr. F. H. Potthoff, III Jack Newman, Esq.
1814 Pine Village Lowenstein, Reis, Newman & Axelrad Houston, Texas 77080 1025 Connecticut Avenue, N.W.
Washington, DC 20037 D. Marrack 420 Mulberry Lane Carro Hinderstein Bellaire, Texas 77401 8739 Link Terrace Houston, Texas 77025
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Texas Public Interest Margaret Bishop Research Group, Inc.
11418 Oak Spring c/o James Scott, Jr., Esq.
Houston, Texas 77043 8302 Albacore Houston, Texas 77074 Brenda A. McCorkle 6140 Darnell Houston, Texas 770/4 J. Morgan Bishop 11418 Oak Spring Mr. Wayne Rentfro Houston, Texas 77043 P.O. Box 1335 Rosenberg, Texas 77471 Stephen 'A. Doggett, Esq.
Polian, Nicholson & Doggett Rosemary N. Lemmer P.O. Box 592 11423 Oak Spring Rosenberg, Texas 77471 Houston, Texas 77043 Bryan L. Baker
,1923 Hawthorne Houston, Texas 77098 Rcbin Griffith Leotis Johnston 1034 Sally Ann 1407 Scenic Ridge Rosenberg, Texas 77471 Houston, Texas 77043 Elinore P. Cummings Atomic Safety and Licensing
- 926 Horace Mann Appeal Board Rosenberg, Texas 77471 U.S. Nuclear Regulatory Commission Washington, DC 20555 Atomic Safety and Licensing
- Board Panel U.S. Nuclear Regulatory Commission Mr. William Perrenod Washington, DC 20555 4070 tierrick Houston, TX 77025 Docketing and Service Section
- Office of the Secretary Carolina Conn U.S. Nuclear Regulatory Commission 1414 Scenic Ridge Washington, DC 20555 Houston, Texas 77043 Mr. William J. Se!messler 5810 Darnell Houston, Texas 77074 The Honorable Ron Waters
. State Representative, District 79 3620 Washington Avenue, No. 362 s '
Houston, TX 77007 w
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Stephen /M. Schinki Counsel for NRC Staff
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