ML19317G654

Forwards Request for Addl Info Re FSAR Sections on Design Bases for Structures & Equipment,Instrumentation & Control Sys,Radwaste & Radiation Protection, & Safety Analysis

ML19317G654
Person / Time
Site:	Rancho Seco
Issue date:	07/19/1972
From:	Deyoung R US ATOMIC ENERGY COMMISSION (AEC)
To:	Davis E SACRAMENTO MUNICIPAL UTILITY DISTRICT
References
NUDOCS 8003250709
Download: ML19317G654 (8)

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n DisrRIButoN 7//9/7p AEG FDK PWR Branch Chiefs Local PDR R. W. Klecker Docket OGC RP Reading Regulatory Operations (3)

PWR-4 Reading B. Buckley S. Hanauer N. Brown (2)

Docket No. 50-312 R. S. Boyd R. C. DeYoung D. Skovholt F. Schroeder R. Mascary Mr. E. K. Davis, General Counsel D. Knuth Saaramente Mtsaisi al UtilitI District P

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Dost Mr. Davis:

.7 On the basis of our continuing review of the Final Safety Analysis Report for the Emacho Sese Nuclear Generating Staties, we find that we need additional information to complete our evaluation. The specific information required is listed in the emelesure.

,j In order to maintain our licensing review schedule we will need a seaplately adequate response by August 21, 1972. Please inform us within seven (7) days after receipt of this letter of your confirmation of the schedule or the date you will be able to meet. If you cannot :cet our specified date or j

if your reply is not fully responsive to our requesta it is highly likely i

that the overall schedule for completing the licensing roview for this projaat will have to be. extended. Einen reassignment of the staff's efforts will require completion of the newf assignannt prior to returning to thia 77

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n REQUEST FOR ADDITIONAL INFORMATION SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION DOCKET No. 50-312 5.0 DESIGN BASES - STRUCTURES AND EQUIPMENT 5.69 The response to Request 5.3 requires further amplifiestion.

5.69.1 Indicate by what method the equivalent static soil pressure acting on the walls has been determined for the seismic design of the structures.

Indicate its numerical relationship to the active and passive soil pressure.

5.70 The response to Request 5.25 requires further amplification.

5.70.1 Provide a sketch showing the differential design pressures, design temperatures, and the value of the jet forces used in the design and impingement area for the different cavities of the interior structure.

Indicate for the most critical load combination the margin of safety and/or the maximum ductility factors for cencrete and reinforcing steel.

5.71 The response to Request 5.46 requires further amplification.

5.71.1 Indicate in detail what additional radial measurement at elevation 65 is planned and explain by what means it can provide a check on out-of-roundness (see Safety Guide No.18). Clarify the contradiction between the text and Figure IA-2 indicating one measurement at elevation 65 and several measurements at elevation 70, respectively.

5.72 Our present thinking on tendon surveillance is as follows:

(a) Number of tendons:

10 - hooo 5 - vertical 6 - done (b) Frequency of inspection:

12, 24, and 36 months af ter the initial containment leak rate test and every 5 years thereafter.

Demonstrate that the surveillance program offered in the FSAR provides the same degree of reliability in detecting an eventual corrosion of the tendons as our recommendation.

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. 5.73 It is not clear f rom Figure 5.2-4 what leak-testing procedure will be followed for penetrations. Indicate:

(a) the part of the penetration that will be shop tested for leaks (b) the procedure to be followed for field-weld leak-testing (c) the method of preventing axial leakage through the cables in electrical penetrations (d) the possibility of testing individual penetrations without pressurizing the containment (e) the procedure for bellows laak-testing i

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m 7.0 INSTRUMEfffATION AND CONTROL SYSTEM _S_

7.24 Supplement your response to Request 7.3 by providing a summary listing (tabulation) correlating all safety related electrical equipment, equipment locations, seismic design bases at each location, and seismic test and/or analysis results. This should include the 4.16Kv switchgear; 480v switchgear and motor control centers; 120v a-c system components; 125v d-c system components including batteries, battery racks, battery chargers, distribution centers,' and panelboards; static inverters;- process control equipment; protection and safeguards actuation racks; nuclear instrumentation; electrical penetration assemblies; etc.

7.25 Supplement your response to Request 7.4 by providing the following additional information which is required for evaluation of your de-sign with respect to the independence of redundant protection systems:

4 (a) Cable derating. The IPCEA-IEEE criteria used should be identi-i fied by title and publication number, and an explicit description of the application of the criteria should be provided.

(b)

In a recently approved plant, fire barriers were required above power cable trays that were located parallel to and below instru-mentation and control trays. Indicate the incidence of similar situations in your design and your intent with regard to installation of barriers (reference FSAR section 8.2.2.10.H.1),

3 (c) Clarify the distinction between and the separation and inde-pendence of reactor protection system and safety feature actuation system power and control Channels A and B, and Channels C and D (reference FSAR sections 8.2.2.10.H.3 and j

8.2.2.10.H.4).

What is your intent with regard to explicit identification and documentation (at the site) of all non-safety related circuits which are run in safety trays?

. (d) ~ Indicate the maximum height of the ce.ble tray sidewalls. If i

this height exceeds 4' inches, provide the design criteria which preclude damsge to and possible failure of the insulation of the 4

bottom layer of cable due to compaction over the design life of the plant.

7.26 Supplement and revise your response to Request 7.5 regarding environ-

menta1 ' qualification of safety related components located in the primary containment to. include the following:

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-4 (a) Provide a concise statement of the limiting DBA environmental conditions in the containment. This should include temperature, pressure, humidity, radiation, and chemical environment.

(b) State the length of time from occurrence of the DBA that each component is required to operate in order to perform its design safety function. Table 7.1-1 should be expanded to include this information.

(c) Describe the environmental qualification testing performed for each component and identity the anplicable test documentation.

Provide this test documentation if it has not been previously submitted. Your response should (1) state whether the tests were performed on prototype equipment, (2) contain sufficient detail to permit a direct comparison between the test conditions and the limiting DBA conditions (superimposed on normal aging) for all parameters, and (3) discuss the adequacy of the environ-mental qualification for each component.

(d)

If environmental qualification is based (in whole or in part) on analyses or on use of data from tests on other than prototype equipment, describe and justify each instance of the use of these methods and identify the applicable documentation.

7.27 The design and Technical Specification of the motor-operated isolation valves for the core flooding tanks (as described in your response to Request 7.15) do not provide sufficient assurance that these valves will open when required. An acceptable variation of your design would j

include the following features.

(a) Valve position visual indication (open or closed) for each valve which is not dependent on power being available to the valve controller.

(b) Visual and audible alarms for each valve when the valve is not fully open and reactor coolant pressure is above a preset value, j

These alarms shall be actuated by redundant and independent

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valve position sensing circuitry including at least one position

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sensor sensing actual valve position, and by redundant and independent pressure signals.

(c) A Technical Specification requirement that the reactor shall not be made critical or shall be shut down unless the motor

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. operated isolation valve in the discharge line of each core flooding tank is open and the breaker supplying power to the valve operator is locked open and tagged (the original response states only that the reactor shall not be made critical...etc.).

please indicate your plans and schedule to modify the design of the isolation valving to include the preferred features or to conform to other criteria that provide equivalent assurance that these valves will be open when required.

7.28 Your response to Request 7.17 indicated that no interlock is being provided to automatically close the motor-operated isolation valves in the suction lines of the Decay Heat System (DHS) when reactor coolant pressure exceeds a selected fraction of the design pressure of the DHS. The alarm scheme proposed in lieu of this interlock is not regarded as providing an equivalent degree of assurance that these isolation valves will be closed when required to prevent overpressurization of the DHS. Please indicate your plans and schedule to uodify the design of the DilS valving to include the preferred interlock feature or to conform to other criteria that provide an equivalent degree of protection.

7.29 Describe the means by which a control room operator is appraised of a reduction in engineered safety features redundancy on an overall systems basis.

7.30 Figure 6.2-1 shows only one channel of pressure instrumentation and one channel of level instrumentation for each core flooding tank.

Redundant channels of pressure and level instrumentation are provided for each tank in most PWR plant designs. Justify your design and show that it provides an equivalent degree of assurance that the core flooding system will be capable of performing its design safety function when required. Your response should include a description of the indications and alarms available in the control room to appraise the operator of the state of readiness of the core flooding system.

7.31 In Table 8.2-2, verify that the loads given in the " expected load" column are the same as the connected (motor nameplate rating) loads.

If this is not the case, define the " expected loads" explicitly and state the corresponding connected loads.

7.32 Provide a more complete description of the design and operation of the plant alarm and annunciator systems in the control room. Your response should include a discussion of the power supplies, actuation logic, testability, and a statement of your intent with regard to periodic testing of the system while the unit is at power.

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. 4 11.0 RADIATION WASTE AND RADIATION PROTECTION 11.7 Discuss measures taken to preclude explosive hydrogen gas mixtures from accumulating in the waste gas storage tank. Discuss control instrumentation that prevents the compressor from starting in the event of inadvertent combustible gas feed to the compressor.

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m 4 14.0 SAFETY ANALYSIS l

14.2 The discussion of the boron dilution accident, loss of load, dropped control rod, loss of coolant flow and steam line failure in Section 14 of the SAR indicates that interlocks, runback, valve operation, and other actions contro11cd through the Integrated Control System (ICS) are required. Since the ICS is not considered a system required for safety, provide the following:

(a) A list of functions and actions that are controlled through the ICS for these accidents.

Identify the component, such as inter-locks, valves, and runback circuits, involved.

(b) An evaluation of the effects of failure of each item listed in (a)above. For the purpose of this evaluation, assume that the failure or combination of failures occurring in the ICS (or ICS controlled equipment) are those which have the greatest potential for serious consequences.

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