ML19274D803
| ML19274D803 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 02/05/1979 |
| From: | Whittemore F HUNTON & WILLIAMS |
| To: | |
| Shared Package | |
| ML19274D797 | List: |
| References | |
| NUDOCS 7902260167 | |
| Download: ML19274D803 (16) | |
Text
2/5/79 UNITED STATES OF AMERICA NUCLEAR REGULATORY CCMMISSION Before the Atomic Safety and Licensing Board In the Matter of )
) -
LONG ISLAND LIGHTING CCMPANY ) Docke t No. 50-322
)
(Shoreham Nuclear Power Station, )
Unit 1) )
MOTION FOR
SUMMARY
DISPOSITION OF SC CONTENTIONS 4 a( ii) , (iii) & (xvii)
- 1. The following contentions were accepted by the Board only for purposes of discovery because they were insuf-ficiently particularized:1 4a. Intervenors contend that the Applicant and Regulatory Staff have not adequately con-sidered individually a number of generic light water [ reactor] safety issues raised by NRC staff members and applicable to Shoreham in accordance with the backfitting require-ments of 10 CFB Part 5 0.109 and/or the gen-eral design criteria of 10 CFR, Par t 5 0, AO-pendix A. This contention includes . . . I23 the following design features for structures, systems, and components:
ii. Lack of independence on [ s ic] ECCS valves.
iii. Analysis of postulated reactor coolant pump rotor seizure in-c id en ts .
1/ See Board Order of March 8, 1978 at 2; Tr. 72.
2/ This ellipsis represents language deleted by the Board. See Tr. 72-73.
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xvii. Improvement of BWR shutdown reac-tivity performance.
SC's Amended Petition to Intervene at 4-5 (Sept. 16, 1977).
- 2. In order to gain a better understanding of those contentions, the Applicant asked several questions. See Second Set of Applicant's Interrogatories to Suffolk County at 1-3 (Dec. 8, 1977). The County's answers provided some additional understanding of the issues that it is trying to raise in these contentions. See SC's Response to Applicant's Second Set of Interrogatories at 3-4 (Jan. 31, 1978) (SC's Interrogatory Fe-sponse). These answers were largely reiterated in SC's Partic-ularized Contentions at 4-5 (Nov. 30, 1978). These points are addressed below in the paragraphs on the applicable conten-tions.
- 3. For the reasons set out in 9 9 4-6 below, SC conten-tions 4a(ii), (iii) & (xvii), as amplified in SC's Interroga-tory Response, raise no genuine issues of fact. Therefore, they are ripe for summary disposition under 10 CFR S 2.749.
- 4. SC Contention 4a(ii). -- " Lack of independence on
[ sic] ECCS valves." Apparently, this contention was spawned by issue no. in NUREG-0138, "Staf f Discussion of Fif teen Techni-cal Issues Listed in Attachment to November 3, 1976 Memorandum from Director, NRR to NRR Staff." See SC's Interrogatory Fe-sponse at 3-4. This contention raises no genuine issue of fact for the following reasons:
- a. Contrary to SC's allegation, issue no. 2 in NUREG-0138 did not raise a safety concern regarding lack of in-dependence of ECCS valves. Instead, the reference to certain redesign work at two PWR's to achieve independence of the ECCS valve interlocks was just usad for illustrative purposes. At issue was the Staff's concern that, when it required some rede-sign on one plant during that plant's operating license review, the applicants for similar plants of ten did not learn about the Staf f's decision until their operating license review. The late receipt of information on the Staff actions has resulted in some applicants having to redesign and backfit their plants, which is possibly more expensive than incorporating the modifi-cation before the design is finalized. The Staf f summarized this issue as follows:
The NRC should advise applicants of potential design problems that have been identified on similar designs in order to permit them to take appropriate action to avoid redesign at the OL stage and possibly more expensive res-olution of the problem.
NUREG-0138 at 2-1. And the Staff's response:
Although this is not a safety issue, the NPP will develop and implement, subject to budget constraints, procedures for systematically apprising utilities with plants under con-struction of potential design problems iden-tified on similar designs so that they can give consideration to them early in the final design stage of their plants.
Id. at 2-2.
- b. Although SC pointed to nothing allegedly wrong with Shoreham's ECCS valves, it suggested under taking an
analysis of the independence of those valves. SC's Interroga-tory Response at 3; SC's Particularized Contentions at 4-5.
Such an analysis is unnecessary and would be repetitious be-cause those valves were designed and that design was reviewed to ensure compliance with the NBC's independence requirements.
Af fidavit of David J. Robare on 4a(ii) at 19 2-4.
- c. SC alleged that Shoreham's ECCS does not comply with 10 CFR Part 50, Appendix A, Criteria 5, 35, and 37. SC's Interrogatory Response at 1; SC's Particularized Contentions at 4-4. To the contrary, Shoreham's ECCS complies with these cri-teria to the full extent applicable. Criterion 5 applies only to multi-unit stations. Therefore, it is not applicable to Shoreham. Criterion 35 requires that the ECCS be designed such that a single f ailure will not jeopardize the functionability of the system. Compliance with this criterict is demonstrated in the Af fidavit of David J. Robare on 4a(ii) at Si 2-4. C r i-terion 37 requires that the ECCS be designed to permit periodic functional testing both while the plant is operating and during shutdown periods. Shoreham meets this criterion. Id . at 9 5.
- 5. SC Contention 4a(iii) . --
" Analysis of postulated reactor coolant pump rotor seizure incidents." Apparently this contention is based on NUREG-0138, issue no. 5, which has the same title as this contention. In that issue, the Staff raised a concern about the possible ef fects of a simultaneous seizure of a reactor coolant pump rotor and loss of of fsite power.
This contention raises no genuine issue of fact for the
-s-following reasons:
- a. The following Staff response in NUREG-0138 in-dicates resolution of this issue:
For the postulated locked rotor accident, it is most likely that of fsite power will con-tinue to be available. Less probable is the case with offsite power continuing only until such time as the turbine generator sheds it load. The most severe case of instantaneous loss of offsite power is quite unlikely; and, in addition, analysis of this more severe case shows that the results are more severe than for the more realistic cases, but still within 10 CFR 100 guidelines. Therefore, re-quiring an assumption of the instantaneous loss of offsite power concurrent with a locked rotor accident would not provide sig-nificant additional safety margins, and this combination is not considered to be a design basis accident.
NUREG-0138 at 5-2; see cenerally id. a t 5 -3 to 5-6.
- b. An analysis of the transient caused by a seized recirculation pump rotor at Shoreham is described in FSAR S 15A.l.22. This analysis shows that the effects would be mild. Af fidavit of David J. Pobare on 4a(iii) at 5 2. In the very unlikely situation that of f site power were lost at the same time that a recirculation pump rotor seized, the effects of these simultaneous events would not be substantially greater than in the case of just the locked rotor. Moreover, any in-crease in offsite radiation would be well below the limits in 10 CFR Part 100. Id. at 9 3.
- c. Although SC provided no reason to suggest that Shoreham would be susceptible to a simultaneous recirculation
pump rotor seizure and loss of offsite power , it suggested un-dertaking an analysis of these concurrent events. SC's Inter-rogatory Response at 3-4.; SC's Par ticularized Contentions at 4-5 to -6. Such an analysis would serve no useful purpose for the reasons outlined in 5 5.b above.
- d. In addition to SC's allegations based on issue no. 5, the County claimed that Shoreham violates 10 CFR Part 50, Appendix A, Criterion 10. SC's Interrogatory Response at 1; SC's Particularized Contentions at 4-4. Cr iterion 10 re-quires that a reactor's:
protection systems shall be designed with ap-propriate margin to assure that specified ac-ceptable fuel design limits are not exceeded during any condition of normal operation, in-clud ing the effects of anticipated operation-al occurrences.
As indicated in 55 2-3 of the Af fidavit of David J. Robare on 4a(iii), no fuel design limits would be exceeded if a recircu-lation pump rotor seized or if that event occur red simultane-ously with a loss of offsite power. Therefore, the Shoreham design meets Criterion 10.
- 6. SC Contention 4a( xvii) . -- " Impr ovemen t of EWR shu td o wn reactivity performance ." Apparently this contention is based on NUREG-0153,3 issue no. 26, which has the same title as this contention. In that issue, the Staff raised a concern regarding the reduced insertion rate of scram reactivity near 3/ "Staf f Discussion of Twelve Additional Technical Issues Eaised by Responce to November 3, 1976 Memorandum f rom Director , NRR to NRR Staff" (Dec. 1976).
-7 the end of a fuel cycle when the control rods are almost fully withdrawn. This contention raises no genuine issue of fact for the following reasons :
- a. The following Staff response from page 26-2 v:
NUREG-0153 indicates resolution of this. issue:
[T] his problem has been resolved as a safety issue by use of Technical i. cifications which result in limitations un control rod withdrawal and/or powe r level . The limit on withdrawal keeps some control rods partially inserted and thus Unproves effective scram response time. Po we r level restrictions im-prove initial conditions for transients and may be either explicit to meet pressure mar-gin requirements during isolation transients or implicit by having to meet mini aum criti-cal power ratios.
- b. In keeping with the Staff's response quoted above, Shoreham's Technical Specifications will limit control rod withdrawal and/or power level to ensure that the scram re-sponse is sufficiently fast throughout the core life. Affida-vit o f Dav id J . Robare on 4a( xvii) at 9 2.
- c. SC also alleged in regard to this contention that Shoreham did not comply with 10 CFR Part 50, Append ix A ,
Criteria 13, 20, and 21. SC's Interrogatory Response at 2:
SC's Particularized Contentions at 4-4. Contrary to the Coun-ty's assercion, Shoreham's reactor protection system, which controls the rapid insertion of control rods, complies with the criteria cited by the County. See Af fidavit o f Dav id J . Ro b a r e on 4a(xvii) at 9 3.
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- 7. For the above reasons, SC contentions 4a(ii) , (iii)
& ( xvii) raise no genuine issues of fact. Accordingly, under 10 CFR S 2.74 9, they are ripe for summary disposition in f avor of the Applicant. We request that disposition.
Respectfully submitted, LONG ISLAND LIGHTING "C"~ 'iY F. Case Whittemore W. Taylor Reveley, III Hunton & Williams P.O. Box 1L35 Richmond, Virginia 23212 DATED: February 5, 1979
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensino Board In the Matter of )
)
LONG ISLAND LIGHTING COMPANY ) Docket No. 50-322
)
(Shoreham Nuclear Pcwer Station, )
Unit 1) )
AFFIDAVIT OF DAVID J. ROBARE ON 4a(ii)
David J. Robare, bainy duly sworn, states as follows:
- 1. I am Senior Licensing Engineer within the Scfety and Licensing Operation of General Electric Company. A statement of my professional qualifications is attached.
- 2. Emergency core cooling system (ECCS) valve independence is governed by Criterion 35 of 10 CFR Part 50, Appendix A. This criterion requires that a reactor plant's emergency core cooling system (ECCS) be designed with features, such as redundancy and separation, to assure that the system safety function can be accomplished, assuming a single failure. This criterion has been imp!emented by the detailed design requirements contairied in 5 4.2 of the Institute of Electrical and Electronics Engineers, Criteria for Nuclear Power Plant Protection Systems (IEEE-279, 1971). ,S,e e Re g u l . c o rv Guide 1.53.
- 3. IEEE-279, 1971 s 4.2 requires that, when a single failure could compromise the integrity of two independent divisions, either that possibility must be eliminated or a backup must be provided 50 that a single failure would be acceptable. Section 4.2 permits sharing (ccmmonality)
of some valves between ECCS subsystems provided that divisional integrity of the logic and sequencing is maintained. Also, interlocks between different logic divisions are allowed so long as the separation require-ments approved by the NRC are met.
- 4. The design of Shoreham's ECCS complies with Criterion 35 because it meets the requirements in S 4.2 of IEEE-279, 1971. See FSAR at 7.3-65, 7.3-72 to 73, 7.3-85 to 86, 7.3-92 to 93, 7.3-103 to 104.
The independence of the ECCS valves and interlocks was confirmed during the independent internal design review that is required by Engineering Assurance procedures.
- 5. Criterion 37 in 10 CFR Part 50, Appendix A requires that the ECCS be designed to permit periodic functional testing both while the plant is operating and when it is shutdown. Compliance with this criterion is demonstrated for the various ECCS subsystems in 9 f of FSAR SS 7.3.1.1.1 .4.
David J. Robare Subscribed and sworn to before me this /" day of 9: o u.. u,, 1979.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )
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LONG ISLAND LIGHTING COMPANY ) Docket No. 50-322
)
(Shoreham Nuclear Power Station, )
Unit 1) )
AFFIDAVIT OF DAVID J. ROBARE ON 4a(iii)
David J. Robare, being duly sworn, states as follows:
- 1. I am Senior Licensing Engineer within the Safety and Licensing Operation of General Electric Company. A statement of my professional qualifications is attached.
- 2. The effects of a Shoreham recirculation pump rotor seizure (wnen one rotor becomes locked in place and the other pump continues to operate) have been studied. This analysis, which is described in FSAR s 15A.1.22, indicates that the effects would be mild. There would be no significant increase in offsite radiation and there would be no fuel failures. See FSAR $ 15A.1.22.5. Furthermore, no fuel design limits would be exceeded because the Minimum Critical Power Ratio would not decrease significantly. Id. at $ 15A.1.22.3.3. Thus, for this transient, the Shoreham design complies with 10 CFR Part 50, Appendix A, Criterion 10, which requires that no fuel design limits be exceeded during anti-cipated operational occurrences.
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- 3. In the very unlikely situation that offsite power were lost at the same time that the rotor seized in a recirculation pump, see NUREG-0138 at 5-2, then the rotor in the other pump would coast down rather than continue to operate as discussed in 'I 2 above. GE has considered what would be the results of these simultaneous events and concluded that the effects would not be substantially greater than the effects of just the lock d rotor. This led to the following conclusions: (a) that any increase in offsite radiation would be well below the limits of 10 CFR Part 100, (b) that no fuel design limits wculd be exceeded, and (c) that, therefore, a formal analysis of these simultaneous events would serve no useful purpose.
David J. Robare Subscribed and sworn to before me this day of 2 ,. ..s.t.., 1979.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY CCMMISSION Before the Atomic Safety and Licensino Board In 'the Matter of )
)
LONG ISLAND LIGHTING CCMPANY ) Docket No. 50-322
)
(Shoreham Nuclear Power Station, )
Unit 1) )
AFFIDAVIT OF DAVID J. ROBARE ON 4a(xvii)
David J. Robare, being duly sworn, states as follows:
- 1. I am Senior Licensing Engineer within the Safety and Licensing Operation of General Electric Company. A statement of my professional qualifications is attached.
- 2. If certain transients occur at Shoreham, such as those initiated by a turbine trip or generator load rejection when the plant is above 30% power level, the control rods must be inserted rapidly (scrammed) to shut down the reactor. The following procedures and design features act in combination to ensure that the negative reactivity insertion rate is sufficiently high at all times:
- a. Shoreham's Technical Specifications provide for the surveillance of reactivity control systems, the reactor protection system instrumentation, power distribution limits, and the safety / relief valve performance to ensure proper scram response throughout core life.
Shoreham's Proposed Technical Specifications s6 3/4.1 .4.
- b. The reactor protection system inserts the control rods rapidly upon receipt of a signal indicating the beginning of a~ transient.
See 5 3 below.
- c. The effects of the transient are-mitigated by the actuation of the recirculation pump trip system and the automatic lifting of the safety / relief valves. See 4 4 below.
- 3. The Shoreham reactor protection system (RPS) is described in detail in FSAR SS 7.2 and 4.3.2.6.3. The RPS complies with the following criteria from 10 CFR Part 50, Appendix A:
- a. Criterion 13 requires that instrumentation and control be provided to ensure adequate safety during anticipated operational occurrences and any accident conditions. The RPS meets this criterion because (1) all input signals to the RPS are monitored and displayed in the control room, and (2) the RPS provides the control to rapidly scram all rods, if necessary. FSAR 6 7.2.1.1.1; see generally id. at S 7.2.1.
- b. Criterion 20 requires that a reactor protection system be designed to automatically initiate a scram su that the fuel design limits will not be exceeded as a result of anticipated operational occurrences and to maintain safety during an accident. The RPS complies with this criterion because it constantly monitors the appropriate plant variables and will automatically initiate a scram if those variables exceed setpoints that are established to comply with this criterion.
Id. at gg 7.2.1.1.3(1)-(2).
- c. Criterion 21 requires that a reactor protection system have (2) sufficient functional reliability that a single failure or removal of a component (or cnannel) will not result in loss of the protection function and (2) the capability to test channels independently when the reactor is operating. The RPS comolies with this criterion because it is designed with four independent and secarated input channels and four similarly designed output channels. No single failure operator action can prevent a scram. Id. at ss 7.2.1.1.3(5), (7).
Moreover, each channel can be tested independently during plant operation.
Id. at 6 7.2.1.1.3(8).
4 If any of the events noted in S 2 above occurs, certain plant features will mitigate the effects of the resulting transient. The first feature is the recirculation pump trip (RPT) system, which is described in FSAR $$ 7.6.1.4 and 7.6.2.4. When the RPT system receives a signal indicating the initiation 3 one of these transients, it causes the main power to be disconnected fr m both recirculation pump motors.
This results in a rapid core ficw reduction, which (a) keeps the core within the thermal-hydraulic safety limits during the transient,and (b) increases the void content. The higher void content supplements the reactivity reduction caused by scramming the control rods. The effects are also mitigated by the safety / relief valves, which are described in the FSAR in NRC Request and Response 212.51. These valves will lift automatically at predetermined pressures and are sized to ensure that during the transients the primary system pressure will not exceed the requirements of the ASME boiler and pressure vessel code Section III, Nuclear Vessels, which is invoked by 10 CFR S 50.55a(f).
U David J. Rocare Subscribed and sworn to before me this ' J '~ day of R.e--,w ., 1979.
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CUALIFICATIONS OF DAVID J. RCBARE My name is David J. Robare. My business address is General Electric Company, 175 Curtner Avenue, San Jose, Cal-ifornia 95125. I am currently Senior Licensing Engineer on the Shoreham project. As such, I am responsible for all technical support work for GE's licensing interfaces with the NRC, LILCO, and Stone & Webster.
I received a Bachelor of Science degree in Electrical Engineering frcm the University of Massachusetts in 1963. I worked for GE as a design engineer in the Large Generator Motor Operation from 1963 to 1967 and as project manager for the Rolling Mill Drive Systems Operation from 1967 to 1970.
In 1970 I became a project application engineer in GE's Nuclear Instrumentation Department. Then in 1972 I was appointed project manager for control and instrumentation systems at reactor sites. From 1974 to 1975 I was lead appli-cation engineer in the SWR Projects Department. And in 1975 I assumed my current position as Senior Licensing Engineer.
In my current capacity I have been responsible for the licensibility of Shoreham's nuclear safety systems, the LCCA and transient analyses, as well as the technical specifica-tions. I am also responsible for the 1.icensing effort asso-ciated with GE's programs concerning loose parts monitoring, recirculation pump performance under accident conditions, and checking for vibration of reactor internals before plant startup.
I am a licensed professional nuclear engineer in the State of California.