ML19263D357

Discusses Applicant'S Position Re FSAR Open Items for Which Responses Have Already Been Docketed,But for Which Staff Expresses Continuing Concern.Issues Include Environ Qualification of Equipment & Sump Level Sys

ML19263D357
Person / Time
Site:	LaSalle
Issue date:	03/23/1979
From:	Delgeorge L COMMONWEALTH EDISON CO.
To:	Parr O Office of Nuclear Reactor Regulation
References
NUDOCS 7903270494
Download: ML19263D357 (8)

Text

, . -

Commonwealth Edison One First National Plaza, Chicago, Ilknois Address Reply to: Post Office Box 767 Chicago, Illinois 60690 March 23, 1979 Mr. O. D. Parr, Chief Light Water Reactors - Branch 3 Division of Project Management U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

LaSalle County Station Units 1 and 2 NRC Docket Nos. 50-373 and 50-374 References (a) : O. D. Parr letter to L. O. DelGeorge dated October 16, 1978 (Set 1)

(b): O. D. Parr letter to L. O. DelGeorge dated December 21, 1978 (Set 2)

(c) : O. D. Parr letter to L. O. DelGeorge dated January 22, 1979 (Set 3)

(d): O. D. Parr letter to L. O. DelGeorge dated February 5, 1979 (Set 4)

(c) : L. O. DelGeorge letter to O. D. Parr dated March 15, 1979

Dear Mr. Parr:

As you are aware, in order to insure that the NRC Staff review of the LaSalle County safety Analysis proceeds expeditiously, commonwealth Edison has resolved to respond to all Staff "open items" identified in References (a), (b), (c) ard (d) by March 23, 1979. In order to accomplish this objective, major information submittals have been made by way of Amendments 41 (January 1979), 42 (February 1979), 43 (March 1979) and 44 (Draft - March 1979) to the LaSalle FSAR. There have been, in addition, numerous canmunications between this applicant and your Staff in an attempt to clarify and resolve issues for which FSAR text mnendments are inappropriate.

9\

\

7903270 %

Commonwealth Edison NRC Docket Nos. 50-373/374 Mr. O. D. Parr: March 23, 1979 It is the purpose of this transmittal to discuss the applicant's position with respect to items for which, in our view, technically defensible responses have already been docketed but with which the Staff has expressed a continuing concern. It is to be hoped that these concerns will be resolved upon review of the following discussion. In the event they are not, it is the judgement of the applicant that these LaSalle issues must be reviewed in the light of recent decisions made by the Staff on similar dockets. If, for example, the Staff reaches a conclusion inconsistent with that reached as a part of the Hatch 2 or Zimmer safety reviews, the burden must rest with the Staff to justify those inconsistencies. The stability of the licensing process depends to a large degree upon the ability of the individual staff reviewer to reach consistent conclusions when presented similar facts. Although continued improvement in system performance is the objective of each of us, selective application of official Staff positions or the bnposition of guidelines that have as yet not undergone the full scrutiny of the administrative process that results in a.. official staff position, frustrate the regulated industry and jeopardize unjustifiably the availability of needed electric power.

It is with this perspective that we ask that you review the materials which follow. Although it is acknowledged with respect to the following issues that reasonable alternative positions may exist, it is the judgement of the applicant that the LaSalle positions are technically defensible and consistent with positions reviewed and accepted by the NRC Staff on similar applications.

I. Open Item Set 2 - No. 5 (Environmental Qualification of Enuipment)

Apparently, in order to verify that the information requirements (Audit Phase) of SPR 3.11, g III.1 have been met, Question 40.110 was issued on the LaSalle County dockets. The applicant responded in FSAR Amendment 39 (October 1978). A thorough review of that response will indicate that all of the information requested in SRP g III.1 has been provided, i.e.,

equipment identification; equipment location; normal, abnormal and accident conditions; time required to operate; and environmental qualification.

Commonwealti1 Edison NRC Docket Nos. 50-373/374 Mr. O. D. Parr: March 23, 1979 Furthermore, an evaluation of the information presented and referenced in the response to Que stion 40.110 should clearly indicate conformance of those materials to the requirements of IEEE-323 (1971). The qualification records are substantially complete and are available for audit by the NRC. The Standard Review Plan in no way suggests that all these records be submitted to NRC for review, although it is acknowledged that all such records should be available for review. In this regard, detailed information has been provided in response to the limited requests contained in Questions 31.4, 31.5, 31.6, 31.34, 31.65, 31.97, 31.205, 40.66, 40.67, 40.88, 40.95, 40.97 and 40.109.

The much borader request contained in Question 40.110 has a dilatory effect on the Safety Review process, especially at this late stage of the review. It is worth noting that Question 40.110 does not explicitly reject the information provided in response to the multitude of questions listed above; rather, the Staff has asked that similar information be provided for all safety-related equipment in the plant. This request is judged to be unwarranted in light of the SRP 3.11 g 2 direction that the evaluation phase of the review " involves the exercise of engineering judgement."

Moreover, the form and content of the request in Question 40.110, in the applicant's view, arbitrarily imposes criteria contained in IEEE-323 (1974) which conflicts with the implementation guidance described in Regulatory Guide 1.89. In this regard, the review process on LaSalle County must be viewed in light of that on the Zimmer project, for which the environmental qualificcation bases are the same. The information request contained in Question 40.110 was not considered necessary on the Zimmer docket.

We invite the Staff to avail itself of the applicant's offer to audit selected additional records rather than requiring the broad submittal requested. We must also request that this review be limited to a confirmation of conformance to IEEE-323 (1971) especially where RE 1.89 is only effective for plants with construction permits on or after January 1, 1974

Conunonwealth Edison NRC Docket Nos. 50-373/374 Mr. O. D. Parr: March 23, 1979 Also identified as an open issue was the radiation environment assumed. FSAR Chapter 12 identifies radiation areas within the plant for design considerations, for operations and for post-accident operation levels. Comparison of the system qualification level for NSSS equipment '2.6x10 7 rads-integrated) and BOP equipment (2x108 . rads-integratet, clearly establishes the conservatism of the qualification basis. Although the Staff alleges that these values are lower than those reviewed for similar plants, the values are in fact identical to those reviewed and accepted on the Zimmer docket.

II. Open Item Set 2 - No. 8 (Degraded Grid Voltage Protection The NRC Staff has requested that a second level of voltage protection for the on-site power system be provided for LaSalle County or that any non-conformances to the Staff position be supported by detailed technical analyses. In the r esponse to Question 40.102, it was indicated that the equivalent system voltage associated with the minimum allowable on-site power system voltage is 320 kV on the LaSalle 345 kV system auxiliary transformer buses. It was determined that the system voltage conditions were such that the probability of the system voltpge degrading to the In addition, the 320 kV level is in the range of 10-15 to 10- .

worst case studied (which resulted in a minimum system voltage of 331 kV) would require the loss of four (4) generators and six (6) transmission lines. That case had a probability of occurrence of 10-15 In determining the probability of generator outages, Publication No. 77-64 of the Equipment Availability Task Force of the Prime Movers Committee of the Edison Electric Institute entitled, " Equipment Availability for the Ten Year Period 1967-1976" was used. In determining the probability of line outages, 426 line years of 345 kV line experience on the Commonwealth Edison system was used. Conversations with the NRC Staff reviewer brought out the fact that the data base for the probability calculation is,in his view, questionable. However, no specific basis challenging the technical adequacy or statistical relevance of that data was offered.

Although we acknowledge the responsibility of the Staff to review the adequacy of the system in question, the imposition of a requirement for a second level of under-voltage relay protection is judged to be unwarranted. The fact that this requirement was stated as a " position" on the LaSalle docket

~

~ '

Commonwealth Edison NRC Docket Nos. 50-373/374 Mr. O. D. Parr: March 23, 1979 ignores the fact that no formal published NRC technical position has been identified; nor has it been judged necessary on other recently reviewed plants (Hatch 2, Zhmner). Unless a technically sound challenge to the statistical argument raised in defense of the LaSalle relay protection system is made by the Staff, imposition of the second level of relay protection is judged to be unreasonable, and in fact contradicts the position stated in Q40.102 which allowed for justification of deviations based on

" appropriate detailed technical analyses." The burden now iests with the Staff to indicate where and to what extent the LaSalle analysis is inappropriate or lacks sufficient technical detail.

III. Open Item Set 2 - No. 9 (Diesel Generator Margin Tests)

The diesel generator margin tests to which a commitment was made in response to FSAR Question 40.104 will be performed as a part of the LaSalle County initial test program. The results of these tests will be communicated to the NRC as committed as soon as possible after the testing has been completed.

IV. Open Item Set 3 - No. 6 (Sump Level System)

The issue, as it has been defined by the NRC Staff, hinges on the alleged inability of the sump level monitoring and alarm system to withstand a single failure. This issue was raised in FSAR Question 10.11. However, no basis was presented justifying the requirement that this equipment satisfy the single failure criterion. The Standard Review Plan 5.2.5 and 9.3.3 that address the sump systems do not refer explicitly to General Design Criterion 21 which states the single failure criterion. Furthermore, FSAR Question 212.85 which also requested information on sump detection and alarm systems explicitly precludes compliance with the single failure criterion.

The sump monitor and alarm system comprise one component of the LaSalle leak detection system. That leak detection system must be assessed as a whole. As indicated in the response to Questions 10.11, 212.17, 212.18, 212.78, 212.83, 212.84, and 212.85 the system, taken as a whole: (1) is available after a safe shutdown carthquake (Q212.18, Q212.78 and Q212.84) as required by Regulatory Guide 1.43 (2) is capable of detecting an increase in unidentified leakage of 1 gpm within an hour (0212.17) which satisfies the

Commonwealth Edison NRC Docket Nos. 50-373/374 Mr. O. D. Parr: March 23, 1979 sensitivity requirements of Regulatory Guide 1.45, (3) will be tested for operability during plant operation (Tech Spec 3/4.4.3.1) and includes four detection subsystems, 2 of which must be operable during operation (Tech Spec 3/3.4.3.1) in accordance with the requirements of Regulatory Guide 1.45, and (4) leakage limits are specified as five gallons per minute unidentified leakage, averaged over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period with a limit of 25 gallons per minute of total leakage for any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period (Tech Spec 3/3.4.3.2) .

The LaSalle leak detection system has the same design basis as previously licensed for Hatch 2, and accepted ir the Zimmer Safety Evaluation Report. For this reason, and '.n the basis of the technical adequacy of the system as described above, it is judged that this item can be closed.

V. Open Item Set 3 - No. 13 (ISI - Quality Group D)

The reactor building clased cooling water system (RBCCW) which is the subject of this item .s not identified as a containment isolation barrier per se. However, the system is designed to Quality Group D requirement. The ASME Code,Section XI, does not require regular inservice inspection of the piping in this system.

Nevertheless, the LaSalle Pre-Service Inspection Program previously submitted to the NRC (Reference e) does include the examination of four valves and six pipe lines in the subsystem. The hydro-static testing requirement for these lines has been fulfilled. Functional testing of the four isolation valves and two test valves is included in the pre-service program. It is judged that this position resolves the issue identified.

VI. Open Item Set 3 - No. 44 (Post Accident Leak Detection)

The issue requests confirmation of the operability of the post-accident leakage detection system following a seismic event or a loss-of-coolant accident. As previously discussed in Section IV above and in response to FSAR Questions 212.17, 212.18, 212.78 and 212.84. These responses clearly indicate that the LaSalle" post-accident" leak detection system will remain operable following a seismic event or loss of coolant accident. It is judged that the applicant has been responsive to the NRC Staff request for information and that the system conforms to the applicable Regulatory Guide 1.45.

~

Commonwealth Edison NRC Docket Nos. 50-373/374 VII. Open Item Set 4 - No. 9 (Environmental Qualification of Equipment)

See Section I above.

VIII. Open Item Set 4 - No. 13 (Rod Block Monitor System)

The NRC Staff has taken the position that the Rod Block Monitor (RBM) is a " protective" system. Commonwealth Edison has indicated in response to FSAR Questions 31.67, 31.117, 31.150 and 31.157 that such systems are not required for safety.

Specifically, the RBM is not a part of the reactor protection system as defined by GDC 20 (Item 1). The function described in GDC 20 is that of systems such as the APRM. The RBM is a control and interlock system. Therefore, it is an aid to the operator which minimizes the need to resort to a safety system.

The functional design of the LaSalle system is identical to that accepted on the Hatch and Zimmer dockets. Each plant has two RBM channels which receive input signals from a number of LPRM channels. The only difference between the plants is the number of LPRM channels, due to the difference in reactor core size.

Although the intercourse to date between the applicant and the Staff suggests that the philosophical difference of opinion on the RBM system classification may not be resolved, it is the applicant's view that appropriate surveillance and testing of t'..e RBM are provided in the technical specification (3/4.1.4.3) to assure the performance of the system notwithstanding its classifica-tion. Therefore, it is suggested that this issue be reviewed in light of the approach taken on other recent dockets, the functional design of which are identical to LaSalle.

IX. Open Item Set 4 - No. 22 / Classification of Reactor Internals)

The NRC Staff has stated the position in FSAR Question 421.5 that justification is required to exclude reactor internals such as the dryers, steam separators and other reactor components under the LaSalle Quality Assurance Program. The Staff directs our attention to Regulatory Guide 1.29 for guidance.

Commonwealth Edison NRC Docket Nos. 50-373/374 Mr. O. D. Parr: March 23, 1979 No explicit reference is made in RG 1.29 to the class of components addressed in Q 421.5. Moreover, neither Standard Review Plan 3.2.1 nor 3.2.2 which address seismic and quality group classification respectively, identify that class of components as being within the purview of 10 CFR 50, Appendix B.

Furthermore, the subject components have been classified in FSAR Table 3.2-1, and no question relative to the proposed classification has been raised by the NRC branches responsible for that review.

The applicant's position relative to this class of components has been clearly defined in FSAR Table 3.2-1 and Appendix B (RG 1. 29) . Since this position has been reviewed by the responsible NRC branches and no issue has been raised by them, and since a similar classification approach has been accepted on other recent dockets; placement of the subject components under the auspices of the LaSalle QA program is unwarranted.

As has been previously stated, the discussion of these items herein was intended to provide a consolidated review of the applicant and Staff discourse documented in the FSAR.

It is the applicant's view that adequate technical bases exist to close each of the items reviewed. If you have any questions relative to the materials discussed, please direct them to this office.

Very truly yours, L. O. DelGeorge Nuclear Licensing Administrator cc: C. Reed