ML20137Y922

Safety Evaluation Supporting Amend 76 to License DPR-54

ML20137Y922
Person / Time
Site:	Rancho Seco
Issue date:	09/30/1985
From:	Office of Nuclear Reactor Regulation
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ML20133D117	List: ML20133D115 ML20137Y885 ML20137Y922
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NUDOCS 8510080120
Download: ML20137Y922 (13)

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o, UNITED STATES g NUCLEAR REGULATORY COMMISSION 7p WASHINGTON, D. C. 20555

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTIMG AMENDMENT N0. 76 TO FACILITY OPERATING LICENSE NO. OPR-54 SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION DOCKET NO. 50-312

1. INTRODUCTION A. DESCRIPTION OF PROPOSED ACTION The proposed action would amend the Rancho Seco Technical ,

Specifications (TSs) to provide conformance with the Inservice Inspection and Testing requirements set forth Section XI of the ASME Boiler and Pressure Vessel Code and Addenda and in 10 CFR 50.55a(g). The action would also amend the TSs governing the inspection of steam generator tubes to reflect operational experience applicable to steam generators of the type utilized at this facility.

B. BACKGROUND INFORMATION By letter dated April 22, 1976, the Comission requested the Sacramento Municipal Utility District (the licensee) to apply for amendment of the Rancho Seco Nuclear Generating Station (the facility) TSs to provide conformance with the guidance contained in Section XI of the ASME Boiler and Pressure Vessel Code and the requirements established by paragraph 50.55a of Title 10 of the Code of Federal Regulations. By letter dated March 16, 1979, the licensee submitted the requested application for amendment. In submitting this application, the licensee also requested shanges to the TS governing the inservice inspection of steam generator tubes.

As a result of discussions with the NRC staff, the licensee submitted letters dated December 12, 1979, February 19, 1985, and April 24, 1985, which supplemented and revised the original submittal.

II. EVALUATION A. APPLICABLE EDITION OF SECTION XI CODE AND ADDENDA Based on the date of comercial operation of the Rancho Seco facility (April 17,1975), facility license amendment dated May 30, 1978, and the provisions of paragraph (4) (iii) of 10 CFR 50.55a(g), the licensee's Inservice Inspection (ISI) and Inservice Testing (IST) programs must conform to editions and addenda of Section XI no earlier than the 1974 Edition and Addenda through Summer 1975. This review has been based on this edition and these

addenda.

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B. CHANGES IN TECHNICAL SPECIFICATIONS

1. Specification 3.3.1. A.7 This specification prescribes the operability requirements for the Borated Water Storage Tank and specifies when the isolation valves may be inoperable. The licensee proposes to revise the paragraph numbers referenced by this specification to reflect the proposed revision of Sections 4.5.1 and 4.5.2, We conclude these are necessary and appropriate editorial changes and that they are acceptable.

2. Specification 3.3.3 This specification prescribes the actions to be taken prior to initiating maintenance on components of the Emergency Core Cooling, Reactor ' Building Emergency Cooling and Reactor Building Spray Systems. The present specification requires that prior to initiating maintenance on any of the components in these systems, the duplicate (redundant) component must be tested to assure its operability.

The proposed change would eliminate the requirement for an actual test and would substitute a requirement that the redundant component be verified operable by checking that the surveillance test for the component has been successfully completed and will remain in effect for the duration of the maintenance period. The proposed revision also states that inservice testing shall not be performed on any components where the redundant component has been declared inoperable or is out of service for any reason.

The proposed change confonns to current NRC policy regarding such suiveillances as stated in the Basis for Specification 4.0.3 in the Standard Technical Specifications (STS). The policy stems from the fact that testing a component in one system will frequently make the entire system inoperable.

Thus, requiring special testing of the redundant component in one train prior to performing inservice testing of a component in the other train could frequently cause both trains of a e safety system to be inoperable. Similarly, requiring inservice testing of a component in one train while the redundant component in the other train was inoperable could frequently cause both trains to be inoper9ble. To avoid conditions where both trains could be required to be inoperable, we have concluded it is preferable to rely upon the records of satisfactory completion and current status of the required surveillances.

Because the revisions proposed by the licensee conform to the NRC policy in this matter, we find the proposed revision acceptable.

3. Table 4.1-2 This table specifies the type of test and minimum test freouency for testing specified components in safety-related systems. Among the components listed in this table are the Pressurizer code safety valves and the Main Steam safety valves. The present specifications require the setpoints of one Pressurizer code sVety valve and two Main Steam safety valves to be tested each refueling interval. The licensee proposes to delete this requirement and substitute the ASME Section XI requirements for valve testing. This change would be implemented by adding a Footnote No. 3 stating this requirement. Inasmuch as this change implements the Comission's requirements as stated in 10 CFR 50.55a(g), we find this change acceptable.

A clarification of control rod testing requirements requested by the licensee in the submittal dated February 19, 1985, will be addressed in a separate licensing action.

4. Specification 4.2 This section presently addresses the requirements for surveillance and inservice inspection of the Reactor Coolant System only. The licensee proposes to revise this section by broadening its scope to include inservice inspection and testing of all ASME Code Class 1, 2 and 3 systems. The i

licensee also proposes certain editorie rearrangement and renumbering of paragra? phs. Inchanges addition,involving the licensee proposes to delete an obsolete specification (requiring preoperational examination - presently numbered 4.2.3) and move the substance of the specification to the Bases portion of this section where it will serve as background.

As noted above, this section also contains requirements relating to surveillance of the Reactor Coolant System. These requirements pertain to the handling of the irradiation surveillance specimens. The licensee's submittal of February 19, 1985, proposed to make changes to these provisions as well. But, in the April 24, 1984, submittal, these changes were deleted. Therefore, the current TSs for handlir.g the irradiation surveillance specimens remain unchanved. However, for editorial purposes the present specification 4.2.8 has been relocated to specification 4.2.1 Because this proposed change extends the requirements of this section to all Class 1, 2 and 3 systems (not just the Reactor Coolant System), because the other proposed changes are editorial in nature or delete obsolete requirements, and because the changes are consistent with NRC policy and

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requirements, we find the proposed changes to the applicability and objectives and the proposed editorial changes and deletions acceptable. Specific subsections of this specification are addressed below.

5. Specification 4.2.2 As proposed by the licensee, this section is now titled

" Inservice Inspection." The proposed section incorporates the substance of the present inservice inspection requirements, adds the Section XI requirements for Inservice Testing, modifies the current inservice inspection requirements to conform to current NRC policy and adds certain provisions contained in the STS relating Section XI requirements to other TS requirements. The individual subsections are discussed below.

6. Specification 4.2.2.1 The first three paragraphs of the proposed revision to this subsection replace the present paragraph 4.2.2 with substantially the same wording that is contained in STS sections 4.0.5.a. d and e. The only significant difference between the licensee's proposal and the STS wording is the retention by the licensee of the existing phrase "as closely as design permits" in referring to the inservice inspection and testing that will be performed. While this phrase could be interpreted as basing the need for inspection and/or testing on the judgment of the licensee, this is not in fact the '

case. The proposed specification clearly states the program must be performed except where written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i). The licensee's requests for relief.from Code requirements have been evaluated and those which the Comission found justified were approved. These approvals are documented in the Commission's letters to the licensee dated January 28, 1983, and September 25, 1984. Accordingly, the question as to those locations where the design does or does not pemit inspection and/or testing has been determined in writing by these references. Based on this, we conclude the licensee's inclusion of the phrase "as the design permits" has no regulatory significance and is therefore acceptable.

The licensee's proposed specifications do not include STS paragraphs 4.0.5.b and c. These paragraphs, however, only identify the intervals at which tests must be performed.

Because these intervals are also defined in Section XI, the Commission letters cited above and Section 1.9 of the facility TSs, we conclude the omission of paragraphs 4.0.5.b and c. is acceptable.

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1 The licensee has also included a provision which states a component shall not be tested if the redundant component is inoperable or out of service for any reason. As discussed for specification 3.3.3, this provision was included to prevent both trains of a safety system being out of service at the same time in order to satisfy testing requirements. Although this provision requires that the licensee not perform testing under such conditions, it does not relieve the licensee of responsibility for timely completion of the required surveillances. As stated in the basis for this specification, the licensee is expected to establish a test schedule that can accommodate a reasonable number of inoperability events without violating required test frequencies. Accordingly, we find this provision appropriate and acceptable.

7. Specification 4.2.2.2 This specification would replace present specification 4.2.4 which now prescribes the inservice inspection requirements for the reactor coolant pump motor flywheels, and documents the previously approved exceptions to Section XI inspection requirements. The proposed revision retains the inspection requirements for the motor flywheels, but deletes the previous exception approvals, including Table 4.2-2. Because the matter of permissible exceptions to inspection requirements has been more recently addressed by the Comission in the letters cited above, the omitted material is no longer necessary. We therefore conclude the changes proposed by the licensee for this paragraph are appropriate and acceptable.

8. Specification 4.2.2.3 This proposed revision would change present specification 4.2.5 by adding Scction XI as the basis for evaluation of indications and repair of defects. Because the present specificatioh does not identify a basis for performing such evaluation and repair and because Section XI is the appropriate reference, we find the proposed revision acceptable.

9. Specifications 4.2.2.4 and 4.2.2.5 These specifications are numbered 4.2.6 and 4.2.7 in the present TS. The licensee proposes to retain these provisions unchanged in the revised specification except for the paragraph numbers. We conclude this is an editorial change which is appropriate and acceptable.

10. Bases for Specification 4.3 The licensee has proposed certain changes to the Bases for the specifications in section 4.2 to reflect the proposed revision's

e 0 to this section. We have reviewed the proposed changes relating to Specification 4.2.2 (IST and ISI) and find them appropriate and acceptable.

11. Bases for Specification 4.3 This specification prescribes test requirements to be performed after opening the Reactor Coolant System. The basis has been changed to delete reference to Code 831.7 and state that repairs and modifications to the Reactor Coolant System are inspectable and testable under the provisions of the ASME Boiler and Pressure Vessel Code,Section XI. Because this conforms to current regulatory requirements, we conclude this is acceptable.

12. Specification 4.4.1.3 This specification prescribes the requirements for functional testing of the containment isolation valves. The present version of this specification allows the licensee to determine when it is impractical to test these valves. Under the provisions of 10 CFR 50.55a(g), inservice testing must be performed under all circumstances except where the Comission has granted written relief. The licensee proposes to revise this specification to reflect these requirements. Because this change would conform this specification to current regulatory guidance, we conclude the proposed revision is acceptable.

13. Specification 4.5.1.1.A This specification prescribes the requirements for perfonning system level tests of the High Pressure Injection System. In paragraph 1, the licensee proposes to change the phrase "A manual trip signal..." to "A manual initiate signal." Because in either case a manual, as opposed to an automatic, signal would be employed, we conclude this is a minor editorial change and that it is acceptable. Similarly, in paragraph 2, the licensee proposes to limit consideration of valves to power i

~ actuated valves because these are the only type that would be automatically actuated by a system test and for which the change in position could be verified. Because this is an appropriate clarification, we find the proposed change acceptable.

The licensee has also proposed either or both of these changes in the following additional specifications: 4.5.1.1.8, 4.5.1.1.D.2, 4.5.2.1.A.2, 4.5.2.1.B.1 and 4.8.1.B. For the reasons stated above, we conclude these are minor editorial changes that are acceptable.

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14. Specification 4.5.1.1.B This specification prescribes the testing requirements for the Low Pressure Injection System. The licensee has proposed changes to this specification to provide a minor simplification and consistency in fonnat with similar specifications for other systems. Because no requirements are reduced, we conclude these proposed revisions are editorial in nature and acceptable.

15. Specification 4.5.1.1.D This is a new specification proposed by the licensee in order to include system level tests of the Nuclear Service Cooling and Nuclear Service Raw Water Systems in section 4.5.1.1.

This specific change moves the testing of the Nuclear Service Cooling and Nuclear Service Raw Water Systems from section 4.5.1.2 (which deals with component tests) to this section.

We conclude this is an appropriate editorial change and that it is acceptable.

16. Specification 4.5.1.2 This specification prescribes the test requirements for components of the Emergency Core Cooling, the Nuclear Service

Cooling and Nuclear Service Raw. Water Systems. Except for present section 4.5.1.2.D (Nuclear Service Cooling and Nuclear Service Raw Water Systems) which, as noted above, was moved to section 4.5.1.1, the existing sections are replaced by the basic Section XI guidance for inservice testing. This guidance includes specification of comprehensive test requirements, including defined acceptance criteria. This is important in the present case because several of the paragraphs which would be deleted by this proposed revision contain acceptance criteria for the pumps and valves to be tested. Because the Section XI requirements are those required by the Commission's regulations, we find this proposed change acceptable.

The specification of a quarterly testing interval is not a change from the present requirement. It is noted, however, that although the current edition of Section XI only requires pump testing on a quarterly basis, the edition presently applicable to this facility requires monthly testing. This matter was addressed by the NRC staff in its safety evaluation of the licensee's requests for relief (NRC letter dated September 25,1984). In its evaluation of this matter (Relief Request PV-24) the staff concluded that quarterly testing of the pumps covered by this specification in accordance with Section XI requirements was acceptable provided such' testing was supplemented by monthly " starting tests" for certain pumps

which had previously experienced more than one failure to start. These pumps were the dual-drive auxiliary feedwater pump, P-318 and the two Nuclear Service Raw Water pumps, P-472A and P-4728. By letter dated January 25, 1984, the licensee had previously committed to perform such supplemental monthly tests.

Therefore, based on the current Section XI requirements for quarterly pump testing, the licensee's present requirement for quarterly pump testing and the licensee's written connitment to perform supplemental " starting tests" for the indicated pumps, we conclude the proposed revisions to this specification are acceptable.

In addition to the above changes to this specification, the licensee proposes to add a new requirement. This requirement would provide that following inservice testing of pumps and valves, the required flow path must be demonstrated operable by verifying that each valve in the flow path that is not locked in position, is in its normal operating position. The I

proposed specification would also restate the requirements of Section XI with respect to verifying the positions of locked valves. Based on the fact that these are measures that would clearly contribute to the safety of operations, we conclude these proposed revisions are acceptable.

17 Specification 4.5,.2 This specification prescribes the requirements for testing the Reactor Building Cooling Systems. The proposed revisions to this section are all of the types discussed above for specifications 4.5.1.1. and 4.5.1.2. These include clarification that the system tests apply to power-actuated valves, the use of " manual initiate" rather than " manual trip," replacement of present test requirements with Section XI requirements, the specification of quarterly pump tests and the addition of a requirement for flow path verification. For the reasons stated previously for specifications 4.5.1.1 and 4.5.1.2, we conclude these proposed changes are acceptable. >

The licensee also proposed to delete the last sentence of the Bases section of this specification. The proposed deletion is based on the fact the sentence refers to operational tests that were to be performed prior to plant startup. Because the plant has been in operation for approximately ten years, and because during this period these systems have been tested in accordance with TS requirements, we conclude the statement is obsolete and the deletion is acceptable.

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18. Specification 4.6 This specification prescribes the requirements for testing the auxiliary feedwater pumps. The principal change proposed by

' the licensee is the replacement of the present quarterly test requirements and acceptance criteria with the recuirements of i

i Section XI. Other changes include reformatting the specification into separate sections addressing system testing i

and component testing (Sections 4.8.1 and 4.8.2, I

respectively) and the addition of a requirement for flow path verification. Also, the changes include the substitution of l " manual initiate" for " manual trip" and adding the clarification that the tests apply to " power-actuated valves." Present specifications 4.8.3 and 4.8.4 are retained in total as specifications 4.8.1.D and 4.8.1.C. respectively.

It is noted that although the proposed specification would require quarterly testing in accordance with the requirements of Section XI, as discussed in conjunction with the proposed changes to specification 4.5.1.2.A, above, the licensee has also provided a written commitment to perform monthly

" starting tests" of the dual-drive auxiliary feedwater pump, P-318.

j Therefore, because these proposed revisions do not reduce the present testing requirements and because they provide conformance with current regulatury guidance, provide increased assurance of safe operation (flow path verification) or are editorial in nature, we conclude these changes are acceptable.

19. Specification 4.13.8.1

' Specification 4.13 prescribes inspection requirements for high energy lines located outside containment. The licensee l proposes to change the reference in paragraph 4.13.B.1 from a specific reference to the "1972 Winter Addenda of the ASME

' Section XI Code" to a general reference to Section 4.2 of the TSs. This is the section that provides the overall requirement for inservice inspection in terms of the

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requirements of Section XI and 10 CFR 50.55a(g). Because the

' regulations specify the applicable Edition and Addenda of the Code (which will change with time), we conclude the proposed change is appropriate and acceptable.

20. Specification 4.17 1

This specification prescribes the inspection requirements for steam generator tubes. The licensee has proposed a number of I revisions to the existing specifications based on operational experience with steam generators furnished by the NSSS supplier for Rancho Seco (Babcock and Wilcox). Each of these proposed changes is discussed below.

Objective. The objective of this specification is presently stated to be verification of the operability of the steam generators and ensuring the structural integrity of the tubes as a part of the reactor coolant boundary. The licensee proposes to revise the statement to say the objective is to verify the operability and integrity of the steam generator tubing as part of the reactor coolant boundary. Since the only inspections required by this section of the TSs are inspections of the steam generator tubes, we find the licensee's proposed statement does not have any practical effect on the prescriptive portions of these specifications and does more accurately describe the effect of the specified inspections. Accordingly, we conclude this proposed cha :ge in the wording of the objective is acceptable.

4.17.1. This specification precent.ly states each steam ,

generator shall be demnn:trated OPERA 8LE by selecting and inspecting steam generators as specified in Table 4.17-1. The licentec proposed to revise this statement to say " Steam generator tubing shall be demonstrated operable...." Inasmuch as it is the tubing OPERABILITY that is detemined by the inspection, and because the licensee has not proposed to change the contents of Table 4.17-1, we conclude the proposed change is basically editorial in nature. Therefore, because the proposed change would not affect the safety of operations, we conclude it is acceptable.

4.17.2. This specification prescribes, by reference to Table 4.17-2, the minimum sample size to be used in the steam generator inspection and the required action, based on the results of the inspection. The licensee proposes to revise this section by renumbering Table 4.17-2 to 4.17-2A and creating a second table of this type numbered 4.17-28. These changes are based on having two classes of steam generator tubes: nomal tubes which are subject to random failures, and special tubes which operating experience has shown to be particularly susceptible to degradation. Based on these two classes, Table 4.17-2A would apply to the "nomal" tubes, and Table 4.17-2B would apply to the "special" tubes.

These "special" tubes are defined by the licensee in proposed new paragraph 4.17.2.e as " Tubes in specific limited areas which are distinguished by unique operating conditions and/or physical construction (for example, tubes adjacent to the open inspection lane or tubes whose 15th tube support plate is not broached but drilled)...."

The motivation for such a change stems from the observation that the bulk of significant tube degradation typically occurs in predictable localized areas. The present specifications properly require augmented inspection in these areas. This,

of course, increases the probability that degraded and/or defective tubes will be found. The present specifications, however, do not take into account the location where degraded or defective tubes are identified. As a result, augmented inspection of the ' troublesome' areas can, per Table 4.17-2, trigger greatly expanded inspection of the ' normal' as well as the ' troublesome' areas. This, of course, would lead to additional inspection effort and radiation exposure in

' normal' areas that was not warranted by operating experience.

The licensee proposes to address this problem by fomally identifying the two classes of steam generator tubes and optionally treating them as two separate populations. If the licensee desires to exercise this option, the specific limited areas must be defined and 100 percent of the tubes in those areas must be inspected. The number of tubes in this inspection would not count toward meeting the overall minimum sample size required by proposed Table 4.17-2A (which is unchanged from the present Table 4.17-2), but neither would the results of this inspection cause an augmented inspection in areas outside this defined region. Inspections outside the

' defined region' would continue to be required in the same-manner and to the same extent as at present, with the same criteria for augmented inspection. The only differences would be as stated above. If the licensee did not desire to exercise the ' defined region' option, there would be no difference from the present requirements.

As part of the licensee's proposal. Table 4.17-28 has been proposed. This table specifies the sample size, and action required for various inspection results for inspections performed within ' defined regions'. The table also provides the option of inspecting all the tubes in the defined area in either one or both steam generators.

Regarding the acceptability of the licensee's proposal, we note the principal effect of the proposal is to allow the licensee to divide the steam generator tubes into two populations - one within a defined area where 100 percent of the tubes would be inspected, and the other being the balance of the steam generator which would continue to be inspected to the same extent as reouired by the existing rules. As noted above, the motivation for this change stems from the fact that the present specifications require greatly expanded inspection samples if more than a nominal number of defective tubes (as few as one) are identified. In addition, the specifications require that inspections be concentrated in troublesome locations. Thus, if there are known troublesome areas in the steam generators (as there are for 8W steam NUREG-0886, " Steam Generator Tube Experience" generators - see

), inspections must be concentrated in these areas. Identification of

. e defects in these areas then causes greatly expanded inspection samples, not only in the troublesome areas, but also throughout the steam generator as a whole.

We believe that where operating experience indicates the existence of steam generator areas with significantly different degradation rates or mechanisms, it is appropriate 4

to base the inspection of each area on the experience and y

' findings for that area. As a corollary, inspection of an area with one characteristic should not be expanded as a result of inspection findings in an area with differing

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characteristics. This, of course, is what the licensee has proposed. However, for augmented inspection of the

) troublesome or ' defined regions', the licensee proposes not t

partial, but 100 percent inspection. At the same time, all of the present requirements will continue to apply to areas outside the ' defined region'. Also, even though the licensee proposes to inspect 100 percent of the tubes in the ' defined regions', the licensee does not propose to take credit for these tubes in meeting the minimum sample size requirement for overall steam generator inspection. In addition, none of these proposed changes appear to conflict with the guidance i provided in Regulatory Guide 1.83, Rev. 1. " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes."

I Based on the foregoing, we conclude the proposed changes to this specification provide improved technical bases for performing steam generator tube inspections. Because the proposed program concentrates inspection in known problem areas i

while maintaining the present level of random sampling and action requirements in the other regions of the steam l

generators, we conclude the proposed changes to these specifications do not diminish the level of assurance of safe f

operation and therefore are acceptable.

4.17.5. The licensee has proposed two changes to this section which deals with reporting requirements for steam generator inspections. The first change would require the licensee to report the results of steam generator tube inservice i

inspections to the Commission in the Monthly Report for the

'; period in which the inspection was completed, instead of the Annual report. Because this provides more timely reporting of ,

operational information, we conclude this change is acceptable.

l The second proposed change is to revise the reference to the i

administrative reports section of the TSs (Section 6.9). We conclude this is a minor editorial change that is acceptable, i

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j Miscellaneous. The licensee has made additional minor editorial changes involving references to the renumbered Table

] 4.17-2A and the proposed new Table 4.17-28. These changes occur in paragraphs 4.17.3.b and c, and 4.17.4.b. We conclude these changes are appropriate to the proposed revision and are j acceptable.

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Bases. The licensee has provided a proposed revision to the

' basis for this TS. We have reviewed this proposed change and conclude it is consistent with changes proposed for this section of the TSs and is acceptable.

21. Specification 6.9.5 1 ,

1 This specification defines special reports cover,ing various

}' operating activities that must be submitted to the NRC Regional Office. The licensee proposes to add an item 'P' 4

titled " Steam Generator Tube Inspection" to reflect the report i requirements of specification 4.17.5. We find this to be an '

acceptable editorial revision.

c III. ENVIRONMENTAL CONSIDERATION i

This amendment involves changes in surveillance requirements and also relates to changes in recordkeeping, reporting, or administrative i procedures or requirements. We have detennined that the amendment 1

involves no significant increase in the amounts, and no significant  !

! ' change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Comission has previously issued a i

' proposed finding that this amendment involves no significant hazards consideration and there has been no public coment on such finding.

i Accordingly, this amendment categorical exclusion set forth in 10meets the eligibility)

CFR 51.22(c criteria (9) and (10). Pursuant for i to 10 CFR 51.22(b), no environmental impact statement or environmental i

assessment need be prepared in connection with the issuance of this amendment. ,

! IV. CONCLUSION We have concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and

, (2) such activities will be conducted in compliance with the Comission's i

regulations and the issuance of this amendment will not be inimical to  ;

i the comon defense and security or to the health and safety of the public. .

Dated

September 30, 1985 Principal Contributors: G. Zwetzig l, i l  !

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