ML19317F711
| ML19317F711 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 10/04/1974 |
| From: | Schwencer A US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Sampson G TOLEDO EDISON CO. |
| References | |
| NUDOCS 8001230603 | |
| Download: ML19317F711 (50) | |
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'A 007 4 T74 Docket flo. 50-34 D A*%
OO The Toledo Edison Comoany O
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ATTil: !!r. Glenn J. Sampson D
T Vice President, Power k
A 300 Edison Plaza O
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Toledo, Ohio 43652 Gentlemen:
As a result of our continuing review of the Final Safety Analysis Report (FSAR) for the Davis-Besse fluclear Power Station, Unit 1, we have developed a number of requests for additional information and a number of Regulatory positions regarding your design. The requests for additional infonnation, listed in Enclosure flo.1, are based on infomation in the FSAR and its amendments and your responses to our first-round requests for additional information. The Regulatory positions, listed in Enclosure !!o. 2, also are based on infomation in the FSAR and its mendnents and your responses to our first-round requests. 'Je recuest that you amend your FSAR, clearly stating your position regarding comoliance with each of the requirments listed in Enclosure flo. 2.
Ne are ::repared to meet with you to discuss further any of our positions to assure complete understanding of the factors at issue and the bases for our positions.
In order to maintain our licensing schedule, we will need your resconses to Enclosure !!os.1 and 2 by December 6, l'374. If you cannot neet this date, please inform us within seven (7) days after receipt of this letter so that we may revise our scheduling.
If you plan to appeal to Licensing management on any of these positions, please advise us of your intentions within two weeks.
The requests and positions listed in Enclosure flos. 1 and 2 are not necessarily complete, but are fontarded for your urly consideration.
Additional requests and positions will be forwarded as they are develooed I
from our review.
Sincerely, Origfaal Sigud by l
A. Schwencer Chief
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l Light Water Reactors Branch 2-3 Directorate of Licensing Enclo_ sures:._
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e 2-Toledo Edison Company ccs: Donald H. Hauser, Esquira DI IBUTI0ti:
The Cleveland Electric cket Files Illuminating Company AEC PDR P. O. Box 5000, Poom 610 LPDR Cleveland, Ohio 44101 LWR 2-3 Reading VAMoore Gerald Charnoff, Esquire FSchroeder Shaw, Pittman, Potts, Trowbridge AKenneke and Madden DEisenhut 91017th Street, ii. W.
RX1ecker Washington, D. C.
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Leslie Henry, Esquire IAPeltier Fuller, Seney, Henry & Hodge EGoulterne 800 On ns-Illinois Building TR BCs 405 fiadison Avenue LWR BCs Toledo, Ohio 43604 ACRS (16)
Mrs. Evelyn Stebbins, Chairman Coalition for Safe Electric Power 312 Park Building 140 Public Square Cleveland, Ohio 44114 C
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ENCLOSURE 1 REQUESTS FOR ADDITIONAL INFORMATION TOLEDO EDISON COMPAhT DAVIS-BESSE NUCLEAR POER STATION, UNIT 1 DOCKET NO. 50-346 9
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9 2-General Comments:
Your responses to the following requests for additional information are required for us to complete our review.
In cases where the request is for the purpose of clarifying or completing your response to previous first-round requests, the same request number is used to identify the request. In cases where a new area of concern ic involved, a new request number has been assigned to identify the request.
As indicated in our August 6, 1974 letter to you, and your August 19, 1974 response to that letter, Items 1, 3 and 4 are still open items in our first-round review.
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Request No.
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3.3.1 Your response to Request 3.3.1 is limited to the effects of atmospheric pressure differential only.
In Section 3.2.2.2, you stated that the effects of wind and differential pressure are concurrently applied to the shield building and the auxiliary building. In view of the above, indicate if the concurrent loads due to pressure differential, wind and the postulated tornado missiles are applied to other Category I structures such as the intake structure, valve rooms, etc., and if the answer is negative, justify your response.
3.5.1 Your response to Item 3.5.1 is inadequate. Table 10-3 of Revision 4 reveals that using Petry equation with 64% energy absorption, control room and spent fuel pool roof could be pene-trated by the postulated turbine missile.
(Table 10-2.)
Clarify your response.
3.6.6
- Provide the design margin utilized against the formation of a plastic hinge in those piping systems inside containment not directly using the criteria of Regulation Guide 1.46, and in those piping systems outside containment not directly using the criteria of the A. Giambusso letter of December 15, 1972. These l
piping systems are discussed in the FSAR in Sections 3.6.2.2.1 and 3.6.2.2.2, and are those which utilize the " span length" method for postulating pipe break restraint locations.
3.6.7 Provide additional information in justification of the dynamic amplification factors and thrust coefficients used in the analysis to demonstrate that they have been conservatively selected. Section 3.6.2.5.6 indicates that an overall value of 1.67 has been used in the general case inside containment. This value is assumed to be a composite produce of the thrust coefficient the gap rebound effects factor and the dynamic load factor.
In Attachment A of our letter of 8/28/73, we indicated that the thrust coefficient in the case of saturated steam or water should not be less than 1.26 and that the gap rebound factor should be 1.5, unless a lower value was otherwise justified by dynamic analysis. With a dynamic load factor of 2.0, and the above values for che other factors, a composite product value of 3.78 is obtained. Justify the use of 1.67 inside containment and a value which is apparently 1.5 outside containment (as given in Section 3.6.2.5.9 of the FSAR) in lieu of more conserva-tive multiplying factors.
3.6.9 Amplify the design details for the guard pipe provided on page
}-115 of the FSAR by indicating the stress limit applicable to the loading pressure.
Discuss the method of access provided to carry out inservice inspection of the flued head to process pipe welds sh'own on Figures 3-12, 3-13 and 3-15.
Provide the number of openings, size,;and pressure seating details.
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3.7.1 Damping factors shown in Table 3-7 are high when used in con-junction with the design response spectra. Since only the damping i
values associated with a stress level of "0.5 times yield point"
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are used in analyzing the structures, components and equipment, other damping values, i.e., those associated with 0.25 Fy,1.0 Fy and beyond Fy, should be deleted from Table 3-7 to avoid confusion.
The coluan concerning " stress level" should also be deleted.
Your response to Request 3.8.l(b) refers to the structural 3.8.1 capacity due to pressurization only. Indicate if the Category I elements have been analyzed for the maximum predicted pressure concurrently with the other pipe break loads such as pipe whip, jet impinging elements, earthquake, etc., according to the Document "B" and describe the method of analysis used. Further= ore, compare the conservatism of the method used with that contained in a paper by k'illiamson and Alvy of Homes and Narver, Inc.,
entitled, " Impact Effects of Fragments Striking Structural
' Elements", 195 7.
3.8.2 Your response to Request 3.8.2(b) is not acceptable.
Indicate the method by which the crane support girder is connected to the containment vessel. If welding is used, specify the type and l
size of the weld.
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o 3.9.8 Provide summary results of component mathematical analyses covering the not:21, upset, and faulted plant operating conditions.
Include typical and maximum stress, def1setion and fatigue results for representative load tests considered for each operating-condition as appropriate. Provide code or other permitted allow-able values with each reported result. This is applicable to ASME Class 1, 2 and 3 components.
3'.9.9 Provide a diagram and the design details of the main steam isola-tion valves, particularly, a discussion of the valves design adequacy to withstand the loading effects of fast closure.
3.10.1 Amplify the response to Request 3.10.1 to demonstrate that:
(a) The single frequency input tests are adequate in that one of the following conditions was present:
(1) The characteristics of the required input motion indicate that the motion is dominated by one frequency (i.e., by structural filtering effects).
(2) The anticipated response of the equipment is adequately represented by one mode.
(3) The input has sufficient intensity and duration to excite all modes to the required magnitude, such that the testing response spectra will envelope the corresponding response spectra of the individual modes.
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7-(b) Testing in which the input is applied to a single axis only J
is adequate in that it was previously shown that equipment
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response along the vertical direction is not sensitive to i
vibratory input along the horizontal direction and vice versa.
(c) Fixture design used in testing is adequate since the following conditions were present in the test:
(1) Fixture simulated actual service mounting.
(2) Fixture caused no dynamic coupling to the test item.
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5.2.6 The response to Request 5.2.6 regarding the method used by Babcock & Wilcox to conservatively estimate the reference temperature (RTNDT) f RCPB materials is incomplete. We cannot complete our evaluation until the applicant submits'the topical report, BAW-10046, referenced in Section 5.2.3.6 of Revision 3 of the FSAR, or otherwise provides the required information.
5.2.8 (a)
The response to Request 5.2.8 is partially acceptable.
In addition verify that Table 5-13 containing stress limits applicable to pumps and vessels will agree with revised Table Sc'_2 now in the FSAR applicable to piping.
(b) See also Request 3.9.4 which requests that summary results of component mathematical analyses be provided. This request is also applicable to ASME Class 1 components.
(c) See also Request 3.9.2 which seeks a clarification on stress limits and loading combinations.
This request is also applicable to ASME Class 1 components.
,s 8-6.2.1 For the design bas'is loss-of-coolant accident, specify the integrated energy release to the containment up to the end of the initial blowdown phase.
6.2.2 With respect to the main steam line break analysis, discuss possible single failures in the main and auxiliary feedwat,er systems by which additional fluid cou.d be added to the affected steam generator. For example, the failure of isolation valves to close in the main or auxiliary feedwater lines should be considered.
6.2.9 (a) For typical vent flow paths in the reactor cavity and steam generator compart=ents, present the method including the assu=ptions made, of calculating the flow coefficients for the vent flow paths.
Also provide the entrance and exit loss coefficients and ft's for all vent flow paths.
D (b) For the postulated pipe breaks considered in the subcompartment analysis, provide tables of mass and energy release data (lbs/sec and Btu /sec) as functions of time (sec) over the time span of
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interest for subcompartment analysis.
(c) The statement is made in the discussion of the reactor cavity analysis, on page 6-16 of the Davis-Besse FSAR, that the insul-ation was assumed to blow off immediately. Describe in more detail the insulation that is being referred to and discuss the validity of the assumption. Also discuss how other removable vent flow path obstructions, such as sand plugs, were treated in the analysis.
(d) In the discussions of the subcompartment analyses, on page 6-16 of the Davis-Besse FSAR, the statement is made that the calculated pressures are below the maximan allowable. Specify the maximum allowable pressures.
(e) Figure 5-4 shows restraint rings around the hot and cold leg pipes of the reactor coolant system, within the primary shield pipe penetrations. Discuss whether or not the restraint rings were considered in evaluating the vent flow path areas for the reactor cavity analysis.
If they were not considered, redo the analysis.
(f) Describe and discuss the function of the restraint rings shown around the reactor coolant system" pipes, within the primary shield pipe penetrations (see Figure 5-4).
Provide drawings of a restraint ring.
(g) From Figure 5-4, it appears that a limited displacement break or split break could ' occur within a pri=ary shield pipr :.aetratten.
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Provide an an.. lysis of a pipe break within a pipe penetration, and compare the results to the design capability of the primary shield.
6.2.12 As requested previously (see Request 6.2.12), provide a curve of air cooler performance showing energy removal rata as a function of containment atmosphere temperature.
6.2.23 In the response to Request 6.2-23, it is assumed that many isolation valve arrangements and seals and gaskets on airlocks, hatches, and
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e flanges are leaktight. Also from the text, it is difficult to determine which containment penetrations and system lines are actually potential leakage paths which could bypass the volumes treated by the emergency ventilation system (EVS) following a LOCA. Therefore, identify all system lines which penetrate the containment and enter areas not served by the EVS, and penetrations which interface directly with areas not served by the EVS. Discuss the basis for estimating the through-line leakage or leakage past seals and gaskets for each penetration.
Provide a tabulation of the estimated leakage for each potential bypass leakage path, and express the total bypass leakage as a fraction of the contain=ent design leak rate. Estimate the leakage from the shield building annulus and other areas served by the EVS dr:ing the time period following a LOCA when a positive pressure exists in these areas.
6.2.27 Identify all high energy lines that pass through the shield building annulus, and indicate whether or not guardpipes have
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been provided. For the high energy lines that are not prov' ec with guardpipes, provide analyses of postulated pipe breaks.<-
a the annulus. Graphically, show the pressure response of the annulus. Provide tabulations of the mass and energy release data for the postulated pipe breaks. Describe the method of analysis, including the assumptions made regarding heat sinks
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p 6.2.29 The instrumentatio* and setpoints that actuate the purge system valves closed should be identified and justified. Specify the purge valve closure times, including instrument delays.
Provide assurance that the safety features actuation system setpoints will be reached and that containment isolation will occur.
6.2.32 Discuss the essential specifications that were used in the design of the containment vent and purge valves; e.g.,
design temperature, pressure, differential pressure, radiation exposure and dynamic loads. Describe the analysis and/or tests that have been or will be conducted to demonstrate that the valves will operate as specified.
6.2.33 Provide a tabulation of the vent areas between the rooms served by the emergency ventilation system, including the shield building annulus.
6.2.34 Describe the proposed leak test program to measure the fraction of containment leakage that bypasses the shield building annulus and other areas served by the emergency ventilation system.
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6.2.35 Specify the capacities of the containment recirculation system fans.
6.2.36 Provide a curve of the hydrogen concentration in the containment as
. a function of time with one train of the containment hydrogen dilution (CHD) system operating. Provide a curve of the containment pressure as a function of time, and specify the time after CHD system operation that the ILniting containment pressure wculd be reached.
o.2.37 Provide a curve of the hydrogen concentration in the containment as a function of time assuming operation of the hydrogen purge systen. Specifiy the capacit'y of the purge system blower.
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7.1.1 With regards to IEEE Std 336-1971, Installation, Inspection and
- Testing Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generacing Stations, it is not clear from the discussion presented in Sections 7 and 8 whether the requirements of this standard have been or will be met for the installation, inspection and testing procedures for instrumentation, sensing lines, electrical and instru=entation penetrations, cabling and raceways, switchgear and panels.
Supplement the FSAR clearly defining the degree of confor=ance to the requirements of this stand,ard.
7.1.2 Provide the results of a review of your operating, maintenance and testing procedures to determine the extent of usage of jumpers or other temporary forms of bypassing functionIs for operating, testing or =aintenance "of safety-related syste=s.
Identify and justify any cac es where the use of the above methods cannot be avoided. Provide the criteria for any use of jumpers for testing.
u.uormal occurrence reports filed with the Coc=ission involved 7.1.3 aumuuu instrument setpoint drift. A large portion of these problems are credited to insufficilnt margins provided between the setpoint setting and the Technical Specifications limits established by the safety analysis. Provide the design criteria to be used in establishing the required range of instruments used in the reactor pro'tection system, engineered safety feature systems and other safety-related
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systems. In addition, provide the desi:n criteria to be used in
- determining what portion of the range'of an instrument may be used for automatic initiation of a protective function.
Your response should provide assurance that adequate margins are
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provided in establishing these setpoint settings so that exceeding Technical Specification limits due to draft are minimized.
7.3.1 The response :o Request 7.3.1 refers to Table 7.3.1-2 that identifies systems or actuated equipment which cannot be tested routinely during reactor operation and will be tested during each refueling shutdown.
The information contained in the referenced Table identifies only the Containment Vessel Isolation System No. 3, Group 2.
Supple =ent ycur response to include the following:
l.
Provide a description of this system and identify the specific components that can only be tested during each refueling shutdown.
2.
Verify that the above system is the only system that cannot be
- tested during reactor operation or supplement your response to include any other system (s) that is (are) applicable.
3.
Clearly identify the "certain equipment that will be tested only when the reactor is shutdown" referenced in Section 7.3.2.5.
I 7.4.1
,The response to Request 7.4.1 which states that taking exception to IEEE Std 279-1971 in the design of the automatic initiation portion l
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of the Auxiliary Feedwater Sy' stem is justified, is unacceptable.
Supplement your response to address Item c(1) in sufficient detail to demonstrate that the automatic control portion of the system (part of the ICS) is not required for safety, and that the operator has sufficient time to perform the necessary manual functions. Your response should also demonstrate that in the event the operator fails to take action (when required) or executes the procedure incorrectly, there is adequate time available to facilitate corrective action to preclude conditions dee=ed unacceptable for plant safety. If your present design cannot be justified, we will require that the design be modified to conform to the staff's requirements as stated in Request 7.4.1.
7.4.2 The response to Request 7.4.2 is in conflict whih FSAR Section 7.4.1.6.1.
Clearly define the degree of confor=ance to Design Criterion #19 of 10 CFR 50, Appendix A.
Describe the equipment and their location outside the control room that are provided for accomplishing prompt hot shutdown of the reactor.
7.6.1 Technical Specification Section 4.11 for Main Steam Isolation Valve (MSIV) Testing provides insufficient information on the testing procedure.to enable evaluatior ef the adequacy of your design. A i
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'recent study published by the Office of Operation Evaluation surveyed valved malfunctions and recommends the following testing requirements l
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a.
Each MSIV on all water reactors shall be closure tested and actuation timed at the cold refueling shutdown.
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Each MSIV on a PWR, where 4 limitation on differential pressure exists for valve reopening, shall be full closure tested and actuation timed quarterly, at the hot shutdown decay heat removal steam flow condition.
c.
Each MSIV on a PWR, whose reopening is not prevented by differ-ential pressure limitations, shall be partial closure tested and cycled quarterly at reduced steam flow conditions.
(Partial closure testing and cycling in the above context implies a valve travel of 10 to 20 percent from the full open position.)
Clearly define the degree of conformance to the above requirements and justify any exceptions you may have on some other defined basis.
7.6.2 The description for E=ergency Diesel Generator Component Cooling Water System (Section 7.6.2.4), :!ain Steam Lin'e Rupture Control System (Section 7.6.2.5) and :!ain Feedwater Line Rupture Control System (Section 7.6.2.6) is inadequate to facilitate a review. The staff's position is that the instrumentation and controls provided should be
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uusA3acu wo meet the requirements of IEEE Std 279-1971 including the requirements for automatic and manual initiation of protective actions at systems level.
Discuss how your design will comply with this position or justify your design on some other defined basis. Supplement your response to include the following:
1.
Provide functional logic drawings for each system to include the sensing equipment through to the actuators, i.e.,
- valves, pumps, etc.
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Clearly identify the interlocks used (if any) and the setpoints utilized for their control.
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Identify the location of these control systems.
4.
Clearly identify the degree of conformance of your design to IEEE Std 279-1971 and justify any exceptions you may have.on some other defined basis.
(Statements of meeting the intent of IEEE Std 279-1971 are too vague and, therefore, inadequate.)
In the event that your design takes exception to the staff's position as stated above, regarding automatic initiation, justify your design by providing an analysis that:
a) verifies that auto =atic initiation is not required for safety, b) identifies the means by which the operator is infor=ed that the automatic initiation portion is inoperative and that manual control is required, and c) describes the operations performed by the operator (include the time required and available) to manually initiate the systems in question. The manual procedures should be of such simplicity as to provide a high degree of assurance that the operator will perform correctly all actions that are necessary to protect the health and safety of the public.
8.1.1 It is not apparent from the information presented in the FSAR what the design criteria is regarding motors of safety-related motor operated valves. Supplement the FSAR to include this criteria and describe how the design of these motors satisfies the following Regulatory position on " Acceptable Design Criteria for Thermal Overload Protection for Motors of Motor Operated Valves".
Q 1.
Thermal overload protection, if provided for safety-related system motor operated valves, shall have the trip setpoint set at a value high enough to prevent spurious trips due to design inaccuracies, trip setpoint drif t, or variations in the ambient temperature at'the installed location. The trip set-point chosen shall be consistant with that of any branch circuit protective device used.
OR 2.
Thermal overload protection may be bypassed under accident conditions and the bypass circuitry shall be designed to IEEE Std 279-1971 criteria as appropriate for the rest of the safety-related systems.
8.1.2 The reco==endations of the recently issued Regulatory Guide 1.75,
." Physical Independence of Electric Systems" are not included in the present system description or design bases. We will require that you clearly identify the degree of conformance of your design with Regulatory Guide 1.75.
Where less stringent criteria are proposed, discuss the reasons for concluding that the less stringent criteria are adequate.
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. 8.3.10 The FSAR states that the fuel oil storage is common to the diesel generators and auxiliary boiler. Discuss the precautionary measures taken to assure the suitability and quality of the fuel for reliable diesel generator operation. Include the fuel oil impurity limits as well as diesel index number or its equivalent and compare the above limits to the recommendations of the diesel engine manufacturer.
9.1.4 (a) Your response to our Request 9.1.4(a)' and
. company ng diagram i
Figure 9.1.4-1, Revision 3, dated November 1973, indicates the path of a Seismic Cateogry 1 water supply to the spent fuel pool. Provide the following:
(1)
A correlation between diagram 9.1.4-1 provided with your Response 9.1.4(a) and Figure 9.1 in the FSAR. Correct
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Figure 9.1 to agree with 9.1.4-1.
(2) Correct Figure 9.1 in FSAR to indicate double valving between the spent fuci pool and normal demineralized water makeup (Line 2 HCC9).
The spent fuel pool demineralizer and filter must be capable to provide adequate purification of the pool water to permit access to the working area and to maintain pool water optical clarity.
However, the demineralizer is sized to remove approximately 50%
of the fission products contained in the spent fuel pool water in 34 hrs. Clarify whether the capacity is based on 1/3 or 1-1/3 core fuel. Also clarify whether the radioactivity level m,
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calculation of the refueling canal water upstream and down-I stream of the demineralizer is based on the higher loading g
of core fuel assemblies.
i 9.1.4 (b) Your response to our Request 9.1.4(b) states thac with a 6
maximum spent fuel decay heat rate of 10.5 x 10 Beu/hr, it will take 19 hrs, for the fuel pool to reach 200 F.
Provide the results of an evaluation of the spent fuel pool evaporation loss (including radiological emissions) Juring the above heat load conditions.
Assume no other cocling methods available.
Include in the discussion, the nu=ber of fuel clad defects that can be sustained.
9.2.2 Figure 9-2, indicates a single 30" service water return header to the ultimate heat sink. Provide the results of an analysis that demonstrates that the 30" service water return header cannot be partially or totally plugged to prevent the service water system from delivering the required flow rate to all safety-related equipment required for normal shutdown or LOCA condition.
9.2.5 The FSAR states both decay heat removal system coolers and pumps are required to cool reactor coolant temperatures from 280 F to 140 F in 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />. What is the tiae requirement for one pump and cooler? The decay heat removal system is also used to maintain reactor coolant and fuel transfer canal refueling water te=peratures
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at 140 F during refueling operations. What are the requirements for pumps and coolers under these conditions? Assuming only one cooler-one pump combination is available, determine the temperature of the reactor coolant and the refueling water and describe how it would affect the refueling equipment being used for refueling operations.
9.3.4 Your response to Request 9.3.4 is not complete. In addition to the information provided. identify all potential sources in the drainage system where a single f ailu.e could cause flooding of areas or compart=ents housing safety-related equipment and, in this event, what affect it would have on the safe shutdown of the plant. Discuss the precautions taken to prevent flooding by the above mentioned Identify the means provided by which the operator will be sources.
alerted that water is entering the area, rocm or compartsent, and methods available for corrective action. Also discuss the capability of the sumps and su=p pumps s_erving these areas so that plant shutdown is not compromised.
9.3.6 Your response to our Request 9.3.6 conflicts with your response to our Request 9.2.3.. Your response to 9.2.3 states a loss of component cooling water to the RC pump would have.no effect on the pump since the pump is designed to operate indefinitely with the loss of the component cooling water as long as the seal injection water is maintained. Your response to our Request 9.3.6 states the RC pump s al injectidtor can operate indefinitely with the loss of either e
component cooling water. Correct the discrepancy, and pro ide the v
s justification for the statement that the RC pumps can operate indefinitely without cooling water.
9.5.1 Your response to our Request 9.5.1 is not complete. Describe the potential fire hazards in each plant area, room or compartment and discuss the fire risk evaluation utilized in the design of the fire detection and protection system.
9.5.6 Your response to Request 9.5.6 is not complete. Refer to Figure 9.5.6-2 accompanying your response and discuss the provisions made in your design of the diesel fire pump room with respect to flooding particularly at the pu=p flange connection at elevation 576'6" when PMF flood water is at elevation 584'-0".
9.5.9 Describe the design of the diesel combustion air intakes and exhaust systems in the FSAR.
Include in your discussion, the precautionary measures taken to obtain assurance that oxygen content of the incoming combustion air will not under any meteorological and accident conditions be diluted.to an extent as to prevent the diesel I
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from developing full rated power or cause engine shutdown. Also discuss the potential of fire extinguishing gaseous medium or noxious gases being drawn into the combustion air system of the diesel generators', thereby degrading their performance or possibly resulting in loss of emergency generators and loss of emergency powe,r supply.
10.3.6 Provide a failure mode and effects analysis which demonstrates the capability of the two 100% capacity steam driven auxiliary feed pumps to maintain safe plant shutdown under conditions consider-ing the high energy line criteria, i.e.,
single active failure of turbine and a passive failure of steam supply line or failure of the auxiliary feedwater inlet line.
Provide the modifications necessary in' order to meet these criterion if the failure analysis indicates that adequate auxiliary feedwater will not be available to the intact steam generator.
12.2.2 In the response to the location of the upstream radiation monitor sampling heads in the ventilation system (as indicated in Request
~ 12.2.2ofRevision6, June 1934,Page12.2.2-1), Figures 12.2.2-1 and 12.2.2-2 have been referenced by the applicant.
These figures are unacceptable to the staff since they do not clearly indicate where the sampling heads are located. The figures show a myriad of ducts, associated equipment and descriptive details.
The staff would prefer a simplified drawing similar to the flow diagram of the auxiliary building ventilation system in Duke Project 81, Docket 50-488 PSAR, Figure 9.4.2-1, which clearly shows the upstream and downstream locations of the ventilation radiation monitors.
D 12.3.4 Identify personnel who maintains and calibrates the portable radiation survey meters. Specify how recordo will be kept for instrument calibration and maintenance.
13.3.2.4 State whether or not an arrangement in writing has been reached (1/18/73) with the Ottawa County Sheriff's Department, and, if so, provide a copy in Appendix C of the Emergency Plan.
15.3.3 The effect of liquid radwaste component failure is missing.
Provide an analysis indicating the radionuclide concentrations which could occur in both 1) the nearest potable water supply, and 2) the nearest surface water in an unrestricted area as a result of leakage based on single failures of components located outside reactor containment containing radioactive liquids. Assume 1% of the operating fission product inventory is released to the primary coolant, failed tanks release 80% of their design capacity, and all liquids from failed components enter the groundwater; i.e., do not assume liquids are retained by building foundations.
Credit for radionuclide removal by the plant process systems, consistent with the decontamination factors in WASH-1258 should be assumed. List all parameters and provide justification for the values assumed in your calculations, including liquid dispersion and transit time based on distance, the hydraulic gradient, permeability and effective porosity of the soil, and the assumed decontamination due to ion exchange by the soil.
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ENCLOSURE 2 REGULATORY POSITIONS TOLEDO EDISON COMPN.7 DAVIS-BESSE NUCLE.\\R POIER STATION, UNIT 1 DOCKET NO.30-346
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General Consnents:
Tour responses to the following Regulatory positions are required for us to compiece our review. The first two digits in the posicion number designate the FSAR section and subsection and the third designates the position number related to that area.
l d
d Position No.
3.2.2 In Figure 9-21 of the Makeup and Purification System, the classifi-cation change of the letdown line from Quality Group B to Quality Group C at valves MU 83, MU 87 and MU 86, and continuing through the purification filter, purification demineralizers, makeup filters, seal return coolers, makeup tank to the suction side of the makeup pumps is not in agreement with current AEC practice. The Regulatory staff position is that the entire letdown line of the Makeup and Purification should be classified Quality Group B.
Since the system is in an advanced stage of completion, demonstrate that with additional nondestructive testing requirements you will impose on systen piping, your proposed classification will provide a quality level essentially equivalent to that associated with Quality Group B.
Indicate those measures that will be taken to imple=ent any additional l
requirements in 1.1 above in ter=s of the quality assurance provisions l
to be applied.
3.6.1 Clarify the response to Request 3.6.3 and Section 3.6.2.5.4 of the l
i FSAR. Our position regarding the consideration of strain rate effects in the design of pipe whip restraints is that an increase in the specified minimum yield strength of not more than 10% may be used in the analysis to account for strain rate effects unless substantiated by applicable experimental results.
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l
_4 3.9.1 The material presented in Section 3.9.1.1 of the FSAR should be expanded to include the following in the piping preoperational vibration and dynamics effects test program:
(a) a list of sei d locations for visual inspection and measurements, (b) the acceptance criteria that will be applied to establish i
that stress and fatigue levels are within design limits.
l 1
(c) corrective action to be taken should the li=its be exceeded.
3.9.2 The response to Request 3.9.2 regarding loading combinations and stress limits is not adequate and should he clarified as follows:
(a) Section 3.9.2.2(a) of the FSAR is assumed to cover the operating basis earthquake and the other loads acting concurrently, including operating transients associated with the upset plant operating condition. The stress level produced by the sum of these loads should not exceed the allowable stress permitted by the ASME Code for the appropriate class of component at the upset plant operating condition.
(b) Section 3.9.2.2(b) of the FSAR is assumed to cover the safe shutdown earthquake and the other loads acting concurrently, including those due to faulted condition events such as pipe break.
The stress level produced by the sum of these loads should not
__. _. 1.
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exceed the allowable stress permitted by the A5ME Code for j
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In the v,
the faulted plant condition for class 1 components.
I case of class 2 and 3 components, the Code does not provide such stress limits except as covered in Code Cases 1606, 1607, 1635 and 1636, and these limits should not be exceeded.
3.9.3 The material which was provided on the operability assurance program for active pumps and valves should be expanded to verify that the program which covers ASME Class 1, 2 and 3 active pu=ps and valves includes the following:
(a) Provision for the testi=g of appurtenances vital to the operation of active pumps and valves whenever documented records of applicable previously conducted tests are not availabla. The testing of such appurtenance should meet IEEEStandard34k.
(b) A requirement that the manufacturer of active pumps and valves stipulates that each pump or valve will operate normally when subjected to the end connection loads associated with the faulted plant condition.
The material in Section 3.10 and in the response to Request 3.10.1 appears to have already covered some of the information requested in (a) above.
. 9._ _
a 7.1.1 The staff has recently identified a concern with regard to the application of the single failure criterion to manually-controlled electrically-operated valves. It has been concluded that where a single failure in an electric system can result in loss of capability to perform a safety function, the effect on public safety must be evaluated. This is necessary regardless of whether the loss of safety function is caused by an active co=ponent failing to per-form a requisite mechanical motion, or by a passive component performing an undesirable mechanical motion. The following staff position presents an acceptable =eans for meeting the single failure criterion with regard to this type of single failure.
1.
Single failures of both active and passive components in the electric systems of valves and other fluid system components should be considered in designing against single failure, even though the fluid system component =ay not be called upon to function in a given safety system operational sequence.
2.
Where it is determined that failure of a single active or passive component in an electric system can cause mechanical motion of a passive component in a fluid system and this motion results in loss of capability to perform the system safety function, it is acceptable, in lieu of design changes that also may be acceptable, to disconnect power to the electric systems of the component. The plant technical specifications should include
a list of all electrically-operated passive valves, and the required positions of these valves, to which the requirement i
for removal of electric power is applied in order to satisfy the single failure criterion.
3.
Electrically-operated valves which are classified as active valves, but which are manually-controlled should be operated from the t_in control room. Such valves may not be included among those valves from which power is re=oved in order to meet the single failur e criterion unless: (a) electric power can be restored to the valves'from the main control coom, (b) valve operation is not necessary for at least 10 minutes following indication of a plant condition requiring such operation, and (c) it is de=enstrated that there is reasonable assurance that all necessary operator actions will be performed within the time shown to be adequate by the analysis. The plant technical specifications should include a list of the required positions on manually-controlled, electrically-operated valves and should identify those valves to which the require =ent for removal of electric power is applied in order to satisfy the single failure criterion.
.4.
When the single failure criterion is satisfied by removal of electric power from passive valves or from active valves meeting the requirements of (3) abov'e, the associated valves should have redundant position indication in the main control
I, room and the position indication system should itself meet the single failure criterion.
5.
The phrase " electrically-operated valves" includes both valves operated directly by an electric desice (e.g., a motor operated valve and a solenoid-operated valve) and those valves operated indirectly by an electric device (e.g., air operated valves whose air supply is controlled by an electric solenoid valve).
Therefore, please provide:
a.
An evaluation of all safety-related fluid systems to identify all valves whose failure can result in the loss of capability to perform a system safety function.
b.
A description of the means provided to meet the single failure criterion in safety-related fluid systems where it is identified that a single failure will result in the loss of capability to perform the system safety function.
c.
If the single failure criterion is satisfied by meeting the requirements of 3 above, your response addressing Item 3(c) should include:
(1) The procedural instructions given the operators for performing the required actions.
(2) The identification of the equipment provided for monitoring the system status and to facilitate operator action.
(3) The definition of plant conditions for which operator action is required.
D 7.1.2 The information presented in the FSAR concerning the battery
' test program is not complete. We require that your design criteria include:
(1) The " Procedure for Battery Capacity Test", as specified in Section 5 of IEEE Std 450-1972, and (2) The frequency of " Performance Discharge Test", as specified in Section 5.3.6 of IEEE Std 308-1971 unless a demonstrable technical basis can be established for a greater interval between perfor=ance tests.
Revise the FSAR to include the above requirements and justify any exceptions taken.
7.6.1 From our review of the Decay Heat Removal (DHR) System (Sections 7.6.1.1, 6.3.2.16 and Figure 6.17), we have concluded that this system is required for safety, i.e., to achieve cold shutdown of the plant. The present design does not meet the single failure criterion with respect va failure (to open) of either of two serially connected isolation valves (DH11 & DH12) in the suction line of the (DHR) pumps.
We will require that the (DER) system design meet the single failure criterion from the standpoint of assuring decay heat removal (i.e.,
cold shutdown) and from the standpoint of precluding overpressuriza-tion of the system, and that the associated instrumentation, control and electrical systems conform with IEEE Std 279-1971 and IEEE Std 308-1971. Therefore, modify your design to meet these requirements,
. or justify the ' resent design on some other defined basis.
p 7.6.3 The design criteria stated in Section 7.6.1.1.2 is incomplete.
The staff's position regarding isolating low pressure system from the high pressure reactor cooling system is as follows:
1.
At least two valves in series shall be provided to isolate any subsystem whenever the primary system is above the pressure rating of the subsystem.
2.
For systems where both valves are motor operated, the valves shall have independent and' diverse interlocks to prevent the valves from being accidentally opened unless the primary system pressure is below the subsystem design pressure.
3.
For those systems where both valves ar a motor coerated, the
~
valves shall also receive a signal to autoeatical'.y close whenever the pri=aty system exceeds the subsystem design pressure.
4.
For those systems where one check valve and one motor operated valve are provided, the motor operated valve shall be interlocked f& event valve opening whenever the primary pressure is above ou the subsystem design pressure, and to autocatically close when-ever the primary system pressure exceeds the subsystem design pressure.
5.
For those systems which are required ^f. T::CS operation, the above requirements are not mandatory "?s !as ayee.e=s will be evaluated on an individual case basis.
6.
Suitable valve position indication should be scovided for the above valves in the control room.
4
Therefore, supplement the FSAR to show full conformance with the staff's position, including the requirements of diverse inter-locks or justify your design on some other defined basis.
8.3.1 Your response to our Request 8.3.l(e) is not adequate. It is our position that engineered safety features such as the diesel generators shall mee.t the requirements of IEEE-308, paragraph 5.2.4. standby power supply subparagraph (6), " Energy Storage".
Provide the design for a Seismic Category I, Quality Group C fuel oil storage and transfer systen to maintain the diesel generctors in operation for a mini =us of 7 days.
9.2.1 Your response to our Request 9.2.6 is not adequate.
In view of the safety-related functions of the CCWS, it is our position that each surge tank (s) shall be provided with a source and make-up system that meets Seismic Cateogy I requirements.
12.2.1 The Davis-Besse program for grab sampling, as indicated by the response to Request 12.2.3,* Revision 6, page 12.2.3-1, is not
~
realistic. To say that "... grab sampling for area surveillance will be monthly" is not defining a meaningful program. The technique of monitoring by grab samples conciders that certain normal operations do not require continuous air monitoring systems to monitor the atmosphere, but only require that intermittent
-samples be taken during the operation to assess evolvement of 1
airborne radioactivity due to inanticipated release. Thus, the grab sampling program should not be related to a fixed routine l
I l
since operations may require grab samples at different intervals as compared to others. The staff, therefore, requires that the program be more flexible and even consider the use of personal air i
8 samplers during certain normal operations to provide more realistic l
close-in breathing samples.
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13.3.1 We do not agree with your position stated in Item e of your response to Request 13.3.5 (1/18/73). If a Site Emergency should degrade into an Offsite Emergency, the alerted offsite support groups would be better prepared to i= mediately provide any =casures required to protect the public.
It is our position that offsite support groups be alerted when a Site Emergency occurs.
16.01 Address Items a(2)(e) and a(3)(g) of Regulatory Guide 1.16,
" Report of Operating Infor=ation". A generic statement such as that given below would be satisfactory:
" Operating information shall be reported in the form and at frequencies specified in USAEC Regulatory Guide No. 1.16 entitled,
' Reporting of Operating Information', dated October 1973."
16.02 A section entitled, " Miscellaneous Radioactive Materials Source",
should be inserted in Davis-Besse Tech Specs.
(This section can have sub-beadings such as " Radioactive Source Leakage Test" or
" Sealed Source Contamination".)
See the following.
SEP 12 %74 l&OP 515 REY. 1 Technical Specification Addition i
Surveillance Requirement Miscellaneous Radioactive Materials Sources Tests for leakage and/or contamination 3,
Source Leakage Test shall be performed by the licensee or -
by other persons specifically authorized Speci fica tion by the Conmission or an agreement State, 35 I IIO*S Radioactive sources shall be leak tested for contamination.
I*
The leakage test shall be capable of detecting the presence c
bje o cor ux con of 0.005 microcurie of radioactive material on the test ifn i d
th a 1
han tiirty a
sample.
If the test reveals the presence of 0.005 microcurfe
{ ateg ep g,
]
e e 9
or more of removable contamination, it shall inmediately be e
s not ex t
xnnh withdrawn from use, deconteiinated, and repaircd, or he 2.
The periodic leak test required does l
P disposed of in accordance with Conmission regulations, ore d ot ng e
sources excepted from this test shall be tested for leakage prior to any Those quantities of by-product n.aterial that exceed the use or transfer to another user unless quantities listed in 10 CFR 30.7) Schedule B are to be leak they have been leak tested within six months prior to the date of use or tested in accordance with the schedule shown in Surveillance transfer.
In the absence of a certificate from a transferor in-Requirements. All other mortes (including alpha emitters) dicating that a test has been made within six months prior to the containing greater than 0.1 r.:icrocuries are also to be leak transfer, scaled sources shall not be put into use until tested, tested in accordance with f f." Wrveillante Requirements.
+.
SEP 12 B74 2
RPOP SIS, REV. 1 Surveillance Requirement (Cont'd) 3.
Startup sources shall be leak tested prior to and following any repair or maintenance and before being subjected to core flux.
y4 B
Bases Ingestion or inhalation of source material may give rise to I
total body or organ irradiation. This specification assures that leakage from radioactive material sources dois not exceed allowable limits.
In the unlikely event that I
those quantities of radioactive by-product materials of interest to this specification which are exempt from leakage
,I testing are ingested or inhaled, they represent less than one :uximum pennissible body burden for total body irradiation. The limits for all other sources (including alpha emitters) are based upon 10 CFR 70.39(c) limits fo.r plot. onium.
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SEP 12 M -
3 RPOP 515. REV. 1 Reporting Requirements Frequency as per Regulatory Guide 1.16 f g Operations Sunanary Results of required leak tests perfonned on sources if the tests reveal the presence of 0.005 microcurie or more of rei..ovatale contamination Records Retention
~ i A complete inventory of radioactive rnaterials in possession shall be maintained current at all g
times.
l Records required to be maintained for five years:
1.
Test results, in units of microcuries, for leak tests perfonned pursuant to Specification Record of annual physical inventory verifying accountability of sources on record.
)
2.
f 5
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16.03 Section 6.7, " Radiation and Respirator Protection Program" is incomplete since the explanation of notes referred to in Table 6.7-1 is missing. The explanation of these notes is below.
ISee the following symbols:
CF: continuous flow D: demand NP: negative pressure (i.e., negative phase during inhalation)
PD: pressure denand (i.e., always positive pressure)
R: recirculating (closed circuit) 2(a) For purposes of this specification the protection factor is a measure of the degree of protection af'orded by a respirator, defined as the ratio of the concentration of airborne radio-active material outside the respiratory protective equipment to that inside the equipment (usually inside the facepiece) under conditions of use.
It is applied to the ambient airborne concentration to estimate the concentration inhaled by the wearer according to the following for=ula:
Ambient Airborne Concentration Concentration inhaled =
Protection ractor (b) The protection factors apply:
(i) only for trained individuals wearing properly fitted respirators used and maintained under supervision in a well-planned respiratory protective program.
(ii) for air-purifying respirators only when high efficiency (above 99.9% removal efficiency by U.S. Bureau of Mines type dioctyl phthalate (COP) test particulate filters and/or'sorbents appropriate to the hazard are used in I
l l
l atmospheres not deficient in oxyg'en.
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(iii) for atmosphere-supplying respirators only when supplied with adequate respirable air.
3Excluding radioactive contaminants that present an absorption or submersion hazard. For tritium oxide, approximately half of the intake occurs by absorption through the skin so that an overall protection factor of not more than approxi=ately 2 is appropriate when atmosphere-supplying respirators are used to protect against tritium oxide. Air-purifying respirators are not raennmanded for 5
use against tritium oxide. See also fcotnote, below, concerning supplied-air suits and hoods.
4Under chin type only. Not reco= mended for use where it might be possible for the ambient airborne concentration to reach instan-taneous values greater than 50 times the pertinent values in Appendix B, Table I, Column 1 of 10 CFR Part 20.
5Appropriate protection factors cust be determined taking account of the design of the suit or hood and its permeability in the contaminant under conditions of use. No protection factor greater ~
than 1.000 shall be used except as authorized by the Co= mission.
6No approval schedules currently available for this equipment.
Equipment must be evaluated by testing or on basis of available test information.
7Only for shaven faces.
NOTE 1: Protection fa.ti.rs for respirators, as may be approved by the U.S.. Bureau of Mines according to approval schedules i
for respirators to protect against airborne radionuclides.
may be used to the extent that they do not exceed the protection factors listed in this Table. The protection factors in this Table may not be appropriate to circum-stances where chemical or other respiratory hazards exist in addition to radioactive hazards. The selection and use of respirators for such circumstances should take into account approvals of the U.S. Bureau of Mines 1
i
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in accordance with its applicable schedules.
NOTE 2: Radioactive contaminants for which the concentration values in Appendix B, Table 1 of this part are based on internal dose due to inhalation may, in addition, present external exposure hazards to higher concentrations.
Under such circumstances, limitations on occupancy may have to be governed by external dose limits.
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D g.
16.04 The Technical Specification activity limits on the reactor and secondary coolant for the Davis-Besse Nuclear Power Station have b'een calculated by the staff. The reactor coolant noble gas and iodine activities are set by a postulated tube rupture accident and the secondary iodine activitt by a steam line break accident. A loss of off-site power is assumed to occur at the same time.
The Technical Specifications should read as follows:
1.
Maxicun Reactor Ccolant Activity Specification Thespecificactivityofthereactorcoolantduetodoseequivalent I-13L_/ shall not exceed 1.0 Ci/ gram during equilibrium conditions or prior to startup. The iodine activity will be alleued to exceed the equilibrium limit but shall not exceed the maxinun values in-dicated in Figure 1 during 48-hour periods during iodine transients.
Operation uith a reactor coolant activity higher than the equilibrium,
value, however, shall not c::cced 5 of the plant total yearly operating tisc. The total specific activity of the reactor coolant due to all nuclides, excluding tritium, with half-lives of core than 1
15 minutes shall not exceed 90/E uCi/ gram.
If the limits specified above are not satisfied, shutdown preccdurcs shall be initiated immediately and the reactor chall be cooled to 500*F or less within cicht hours after detection.
If the icdine activity exceeds the indicated maximum value the equilibrium iedine --
activity limit must be lewered by an anount proportionate to the amount by which the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> limit was exceeded.
Sampling Frecuency Whenever the reactse is critical, the sampling frequencies given in Table 4.2.1 shall be used to determine prinary and secondary coolant 1/ Dose equivalent I-131 concentration is defined as that concentration of I-131 which alone would prcduce the same dose as the quantity i
and isotopic mixture actually present.
.i
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_FICURE 1 FRACTION (F) 0F DOSE EQUIVALENT I-131 REACTOR C00LCT SPECIFIC ACTIVITY EQUILI3aIUM LIMITS FOLLOWI:iG FO:lER TRJu!SIE::TS FOR DAVIS-BESSE IiUCLEAR POWER STATION hiaxi=u= A11ottaole Concentration = F x Equilibrium Activity) ii I
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250 l
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20 30 40 50 60 70 80 90 100 Resulting Reactor Power After Pcwcr Transient
(% of Rated Power)
5
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activity levels.
In addition, the sampling program described in Table 4.2.2 shall be initiated,
(a) following a change in power exceeding 15 percent of rated power which occurs within one hour period; (b) during the first 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following a partial or complete depressurization of the reactor coolant
-system; and (c) whenever the reactor or secondary coolant activity exceeds the equilibrium limits specified above.
These tests shall be recorded together with the follcwing information:
(a) reactor power history starting 48 hcurs prior to the first sample, (b) fuel burnup by core region, (c) clean-up flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample, and (d) history of de-gassing cperations if any.
Rooortine Requirements _
The above information shall be included in the semi-annual operating report.
If the limits of the specification are exceeded, a report shall be made to the Directorate of Licensing within 30 days.
TABLE 4.2.2 Sampic Measurements Frecuenev Reactor Coolant Iodine activity, isotepic analysis Every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
- liquid sample including at least I-131, I-133, I-135 Basis The basis for the reactor coolant iodine activity limit is a comouted dose to the thyroid at the site boundary of 1.5 rem during the 2-hour period following a steam generator tube rupture accident and a RC to secondary steam generator leakage of 1.0 gpm.
- In all cases, the minimum number of samples shall rot be less than three
- for each transient.
6
22 -
The basis for the total activity limit is a computed whole body dose at the site boundary of 0.5 rem during the 2-he ar period following a steam generator tube rupture accident. Fifteen *2inutes is the average tL=e it
, takes the released gases to reach the site boundary under adverse meteorological conditions and E is the average energy of betas and gammas per disintegration.
The allowance to exceed iodine equilibrium activity 11mits during the specified 48-hour periods following initiation of shutdown or a power transient is made in order to account for possible iodine spiking phenomet.a. The basis for the allowable maximum values is a resulting dose to the thyroid at the site boundary of 30 rem during the 2-hour period following a steam generator tube rupture accident.
~4 3
In calculating the limits specified above a X/Q value of 5.2 x 10 sec/m and an iodine decontamination factor of 10 between the stean and water phases were used.
2.
Secondarv Svsten Activity Specifications The specific activity of the secondary coolant due to dose equiva-lent I-131 shall not exceed 0.05 uCi/; ram.
If.the limits sp;cified above are not satisfied, shutdown proccdures shall be initiated within four hours and the reactor shall be cooled to SC0*F or less within eight hours after detection.
Basis JIhe basis for this specified limit is a computed dose to the thyroid at the site boundary of 1.5 rem during the 2-hour pericd following a main steam line break accident, This is also based on a RC to secondary steam generator leakage of 1.0 gps.
No specifications are made in regard :o noble gas' activity because these gases are assumed to be continuously removed through the condenser.
In calculating the limits specified above a X/Q value of 5.2 x 10 sec/m3 and an iodine decontamination factor of 10 4
between the steam and water phases in the unaffected steam generator.
9
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e
... s 23 -
ITEM 1 of TABLE 4.2-1 MQUENCIES FOR SAMPLING TESTS Maximum Time l
Between Check Frequency Tests i
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1.
Reactor Coolant Cross Activity 3 Times / Week 4 Days Liquid Samples Isotopic analysis to determine equivalene I-131 concentration Bi-weekly 15 Days R_adiochemical for Semi-annual (2) 7 Months E Determination Tritium Activity Weekly 10 Days II)When radioactivity level exceeds 25 percent of limits in specifications 3.1.D, the sampling frequency shall be increased to a minimum of once each day.
I2) Redetermined if the reactor coolant radioactivity changes by more than 10 pCi/cc in accordance with specification 3.1.D.
1 8
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I 17.2.1 We require TE to show the Station Review Board and the Company i
Nuclear Review Board on Figure 17-2 and indicate the reporting l
r i
levels of both of these boards.
17.2.2 We require TE to identify the management position responsible
{
I for the final review and approval of the Davis-Besse QA Program and QA Manual.
17.2.3 We require TE to describe those p.rovisions established for communicating to all responsible organizations and individuals that quality policies, canuals, and procedures are mandatory 1
requirements which must be implemented and enforced.
17.2.4 We require TE to demonstrate the framework for the implementaion of 10 CFR 50, Appendix B, by cross referencing each QA procedure in Table 17-4 to the criteria of 10 CFR 50, Appendix 3.
17.2.5 We require TE to identify the individuals or groups responsible i
for reviewing and approving prior to use those QA/QC documents and procedures addressed in Section 17.2.
17.2.6 TE has not adequately described the independence requirements imposed on inspection personnel. We require TE to fully describe the independence requirements imposed on those individuals or groups performing acceptance inspection or-verification on safety-related activities associated with major maintenance, modifications, and repairs and identify the group primarily responsible for performing this acceptance inspection.
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