ML19331B230
| ML19331B230 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 08/30/1978 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | Howell S CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| Shared Package | |
| ML19331B231 | List: |
| References | |
| RTR-REGGD-01.XXX, RTR-REGGD-1.XXX NUDOCS 8007280823 | |
| Download: ML19331B230 (36) | |
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Docket Nos: C50-3293 50 330 Coasumers Power Company ATTN:
Mr. S. H. Howell THIS 0O,CUMENT CONTAINS Vice President 212 West Michigan Avenue P00R -QUALITY PAGES Jackson, Michigan 49201
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Gentlemen:
SUBJECT:
SUPPLEMENTAL REQUESTS FOR ADDITIONAL INFORMATION:
PART 1 In continuing our review of the PSAR for Midland Plant i
Units 1 4 2, we find that insufficient information has l
been provided for our review to proceed with development i
of staff positions which had bean scheduled for issuance at this time.
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3.i Supplemental requests for information which we require y
for developing our positions are listed in Enclosure 1.
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. is not complete as several Technical Review
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Branches have found it necessary to await receipt of I
additional information before continuing their review.
l We anticipate a second set of supplemental requests will i
be issued in early September 1978.
Additionally, t lists our previous requests for which your reply is acknowledged to be incomplete or preliminary.~
We are presently assessing the status of our review in conjunction with some existing limitations on staff man-power resources.
Our revised schedule for Midland will
-be established following our scheduled meeting with you i
j on August 31, 1978.
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Our letter of April 21, 1978 forwarded a request inadvertently numbered 110.30, rather than the intended 110.31.
We would l
l appreciate your correction as part of your response.
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-.a Please contact us if you desire clarification or other i
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Sincerely, Original Signed byg
- 3. A. Yarga Steven A. Varga, Chief Light Water Reactors Branch No. 4 Division df Project Management
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,m 010-1 010.0 AUXILIARY SYSTSMS BRANCH 010.39 Your response to our request 010.3 states that portions ffj of the auxiliary building roof are not pru*ected from 5
It is our pos tion that any safety related equipment, including new and spent fuel be protected from tornado generated missiles.
Demonstrate that all such equipment located in the unprotected pcrtion of the auxiliary building will not be subj ect to tornado missile damage.
Also identify the equipment that could be damaged due to tornado missiles through the unprotected portion of the auxiliary building roof.
010.40 Your response to our request 010.3 shows that the. auxiliary (3.5) building HVAC exhaust stacks are not protected from tornado -
missiles.
Demonstrate that failure of these stacks due to tornado missiles will not result in damage to safety related equipment or preclude the operation of any safety related ventilation systems.
010.41 You state that seismic events are not considered concurrent (3.6)
(RSP) with a non-LOCA pipe break (i.e., breaks other than the reactor coolant system piping).
There is no basis for this assumption for non-seismic piping systems since a seismic event could cause a pipe break in r non-Category I piping system.
It i.c our position that you revise your analysis and design as necessary to withstand a pipe break in non-Category I systems as a result of an SSE with reliance only on seismic Category I systems for safe shutdown and cooldown in accordance with position C.2 of Regulatory Guide 1.29.
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010-2 010.42 FSAR Section 3.6.2.5 states that portions of the auxiliary (3.6) feedwater system suction piping are located in the turbine building.
Indicate how these lines will be protected-in the event of a tornado or seismic event that damages or collapses the turbine building.
010.43 The design of your main steam and process steam piping (3.6)
(RSP) over the auxiliary building roof is based on specific break locations for the non-seismic portions of the steam piping which include about 1800 fe'et of piping for which no breaks are postulated.
The pipe break criteria in Section 3.6.2.1.3 for non-nuclear class piping which we reviewed and approved by letter dated September 24, 1976,.
from S. A. Varca to S. H. Howell, was limited to seismic Category I piping.
It is our position that you design this piping to seismic Category I requirements in accordance with Position C.2 of Regulatory Guide 1.29 or demonstrate that cold shutdown will not be precluded as a result of i
a pipe break anywhere in the non-seismic portion of the main steam and process steam lines.
Specifically, your design should protect the following from pipe whip and jet impingement.
(1)
l (2)
All safety related equipment below the auxiliary building walls; (3)
All safety related equipment within the auxiliary building walls; and
.n 010-3 (4)
All safety related main steam piping which could be damaged by a break in the non-safety related mr. a steam or process steam piping.
010.44 Your response to our request 010.4 states that you take (3.6) some exceptions to Subsections B.2.c and B.2.d of our Branch Technical Position APCSB 3-1.
Identify specifically what exceptions are taken to these subsections and provide justification for your alternat'ive criteria.
010.45 FSAR Section 9.1.1.3 states that the new fuel racks are (9.1.1)
(RSP) not designed to withstand the maximum uplift forces exerted by the fuel handling system since all hookups to the new fuel assemblies are done manually and under administrative controls.
This does not provide protection against a fuel assembly being struck in the fuel rack with resulting excessive force-being exerted by the fuel handling system.
It is our position that the storage racks and anchorages be able to withstand the maximum uplift forces available from the fuel handling system without an increase in Keff.
Demonstrate how your design meets this position.
010.46 Your response to our request 010.22 is unacceptable.
You (9.1.4)-
.l (RSP) state that the probability of a single failure of the cask crane occuring'while moving over an elevation change is sufficiently small so as not to warrant consideration.
It is our position that the cask shall be prevented from j
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.x 010-4 dropping and tipping into the spent fuel pool.
If your present design does not preclude this accident, than we request that you provide a single failure proof cask handling crane in accordance with our Branch Technical Position ASB 9-1 or Regulatory Guide 1.104.
l 010.47 Your response to our request 010.22 is incomplete.
On (9.1.4, i
1.2) your arrangement drawings in Section 1.2, show the cask travel path in order that we may determine the safety related equipment immediately beneath the cask, or verify that the cask travel is always along the centerline of the auxiliary building.
010.43 Your response to our requests 010.12 and 010.26 are unacceptable.
(9.2.2)
The October 10, 1977 GE report you refer to only addresses the reactor coolant pump motor bearings, not the pump bearings.
The GE report contains only analysis without testing and doe s not show that your reactor coolant pumps can operate for 62 minutes as you have indicated.
Therefore, r
revise your CCW system design, including safety grade instrumentation, to meet the criteria specified in our request 010.42.
Acceptable methods were listed in Approach No. 2 of our request 010.12.
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010-5 010.49 Your response to our request 010.17 states.that the (9.3.1, 10.3) atmospheric dump valves can be man' ally operated by a local (RSP) handwheel.
It is our position that you demonstrate by actual testing, that a controlled cooldown can be safely achieved based on manual operation of the atmospheric dump valves in order to meet our latest requirements that the plant be brought to safe cold shutdown using only safety grade equipment (See also related or request 211.35).
010.50 Your response to our request 010.29, states that when in (9.3.4) the hot idle condition, the reactor coolant pumps can
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withstand damage to the pump seals, provided the seal inj ection or cooling water is restored within 1 minute.
You further state that the makeup / seal injection pumps are reenergized within 15 seconds following loss of offsite power.
Demonstrate that any single active failure, including failure of a diesel to start, or loss of instrument air will not result in loss of-seal injection following the loss of offsite power.
If your design does not meet the single failure criteria following the loss of offsite power, then revise your design as necessary such that a single failure following loss of offsite power will not result in damage to the reactor coolant pressure boundary.
010.51' Your response to our request 010.33 is unacceptable.
You (10.4.9)~
(RSP) state that plant cooldown with the main condenser out of r
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010-6 service is not a design basis for the Midland Plant.
It is our position that the plant should be capable of reaching cold shutdown using only safety grade equipment assuming any single active failure.
Revise your-design as necessary to meet this position (See related request 211.35).
010.52 Your response to our request 010.38 is t'.nacceptable.
(10.4.9)
A single failure in the control circuits (Drawing E-58) to the level control valve 1LV 3875 or 2LV 3875 following a main steam or feedwater line break inside containment could cause the level control valve to the unaffected steam generator to close; this would result in loss of all
. AFW flow to the affected unit since both AFW pumps must use this flow path.
Revise your design such that no single failure will result in loss of all AFW flow following a main steam or feedwater line break.
010.53 In Table 3.2-1 you state that the spent fuel poo.'. liner (9.1.2) plate is not designed to seismic Category I requirements.
Provide the results of analyses which show that a seismic event wil not cause:
(a)
Failure of the liner plate which could affect the ability to adequately cool the spent fuel.
Your analysis should take into account the possibility of flow blockage in the event that portions of one complete section of the liner plate were on top of the fuel racks; and
010-7 (b)
Failure of the liner plate which could result in mechanical damage to any fuel assemblies.
(c)
Leakage from the concrete fuel pool wall that could result in damage to safety related equipment, or uncontrolled release of radioactive fluid to the environs.
Include the assumptions used in estimating the leakage and show how the leakage will be collected and processed.
If the failure of' the liner could result in damage to seismic Category I equipment or to the spent fuel, then provide a l'iner plate in accordance with position C.2 of Regulatory Guide 1.29.
,n 110-1 110.0 MECHANICAL ENGINEERING. BRANCH 110.32 Your responses to questions 110.1 and 110.7 are not acceptable.
(3.6.2)
You have proposed in Revision 8 to the FSAR to implement an (RSP) inservice inspection program on your no-break piping which is in excess of the ASME Section XI (1974 with Summer 75 addenda) requirements. However, your proposed program does not comply with the staff's augmented inspection program described in Standard Review Plan sections 3.6.1 and 6.6.
Addi t,ionally, you have not provided acceptable justification for not imple-menting this augmented inservice inspection program. There-fore, the staff will require that you commit to implement an augmented inservice inspection on your containment penetration piping for which you have used the no-break option. This program shall comply with Standard Review Plan sections 3.6.1 and 6.6.
In sumary, your program must provide for 100% volumetric examination of 100% of the process pipe welds (including redundant trains) in your no-break region during each 10-year inspection interval.
110.33 In FSAR section 3.6.2.1.3 you have provided the criteria by (3.6.2.1.3) which postulated pipe break locations are chosen for non-nuclear (RSP) piping.
For point of clarification, we define non-nuclear to j
mean non-ASME Section III. The term non-nuclear has nothing to do with whether or not a pipe is seismic Category I.
Our review indicates that you have used the criteria of 3.6.2.1.3 for choosing break locations for ncn-seismic j
Category I piping. Some of this piping runs adjacent to the auxiliary building. The failure of this piping appears to have the potential for damaging safety-related equipment.
This may violate the intent of Position C.2 of Regulatory Guide 1.29 and Paragraph B.3.d of Branch Technical Position APCSB 3-1.
The staff requires that you postulate breaks everywhere along the length of non-seismic Category I piping. The choosing of break locations for non-seismic Category I piping based upon a stress of fittings criteria is unacceptable.
110.34 The subject of hydraulic snubbers will remain open until (3.9.4) submission of your response to our question 110.6. You have committed to provide this response in. December 1978.
110.35 The subject of thrust coefficients and finite break opening (3.6.3.2) times will remain open pending our review of topical report
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BAW-10132P. You have committed in your response to 110.9 to revis this topical rt,vt in September 1978 to address these subjects more fully. However, you have stated in the topical report that finite break opening times would be justified on individual plant dockets. We require that you provide the requested information in one place or the other.
110.36 Your response to question 110.15 is not entirely acceptable.
(3. 9. 2.1 )
We will require that as a minimum the following piping systems (RSP) be checked for abnormal transient or steady-state vibration and for restraint of therma. growth. Provide a commitment to test the following:
1.
ASME Code Class 1, 2 and 3 piping systems, 2.
6ther high energy piping systems enclosed in seismic Category I structures.
3.
High energy portions of systems whose failure could reduce the functioning of any seismic Category I plant feature to an unacceptable level, and 4.
Seismic btegory I portions of moderate energy piping systems located outside containment.
We do not require all of these piping systems to be instrumented 1.f visual observation can provide a meaningful test. You should
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provide a list of those piping systems to be tested. You should specify whether the piping system will be instrumented or visually observed. You have committed in 110.15 to provide a description of the acceptance criteria for instrumented lines by September,1978.
Similarly, provide a description of how visual observations will be made, the qualifications of the personnel who will perform these visual observations, and the acceptance criteria they will use, f
110.37 Your response to question 110.17 is not entirely acceptable.
l (3.9.3)
FSAR Table 3.9-3b does not shov LOCA loads as being considered (App. 3A) in the design of these NSSS Class 1, 2 and 3 components.
You (RSP) must demonstrate that these coreponents can perform their safety function (whether active or passive) under the combined loadings of LOCA and SSE. Describe how t;his loading combination has been taken into ac::ount in the design of these components.
110.38 What method have you used in combining the peak responses to (3.9.3) dynamic loads such as OBE, SSE, LOCA, and relief valve discharge? Specifically address the methodology used for response combination.
If the methodology u.ad varies by ASME
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110-3 Code classification or between the NSSS and BOP scope of supply, provide this information for each portion.
110.39 The subject of seismic qualification of appurtenances of (3.9.3) active pumps and valves will remain open until submission (App. 3A) of the qualification summaries.
You have comitted in your response to 110.23 to provide these summaries in August,1978.
110.40 Your response to 110.26 did not specify a date for submitting (3.10)
- the requested information. To expedite the Midland review you should commit to provide this information soon.,
110.41 You have stated in your response to 110.28 that you have (3.9) maintained the primary stresses.belcw yield for those ASME Code Class 2 and 3 piping systems that require functional capability'following an emergency or faulted plant condition.
In order that tne staff may complete its review, provide the following information.
1.
In FSAR table 3A.l.48-4 you have referenced Code Case 1606-1 for Class 2 and 3 piping allowable stresses.
Verify that those Class 2 and 3 piping systems requiring functional capability did not utilize the higher stress limits allowed by this Code Case under emergency and faulted conditions.
2.
Verify that the primary stresses in these systems did not exceed ' yield during as well as following the faultad loads.such as SSE.
110.42 The response to 110.30 is unacceptable. The qualification (6.2.4) summaries for the containment purge system isolation valves (lM0-5309A,B; 2MO-6209A,B; IMO-5316A,B; 2MO-6216A,S) have not been provided as of Revision 10 to the FSAR. When this sumary is provided, it should specifically address the capability of these valves to close against flow resulting from the LOCA pressure and temperature inside containment. Additionally, SSE loadings should be assumed to act on the valve during this closure against LOCA induced flow.
110.43 In FSAR Table 3.9-1 you have shown Code Case 1711 as containing (3.9.3) the stress limits used for various pressure relief valves. We have endorsed this Code Case subject to the following additional information being provided:
If stress limits are used in excess of those specified for the upset operating condition, you must demonstrate how the pressure relief function is assured.
If no pressure relief function is
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110-4 required when higher stress limits are imposed, this should be clarified.
This information could best be presented in the qualifiestion summaries for these valves. These summaries have not been provided as of Revision 10 to the FSAR.
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Provide a schedule for submitting the remaining qualification l1 0.44 (3.9.3) summaries in FSAR Table 3.9-1.
110.45 Your response to 110.13 requires further clarification. The (3.6.2.2) non-linearity of honeycomb force-deformation curves which you discussed in your response to 110.13 was not exactly the non-linearity with which the staff is concerned. The curve below will clarify our concern.
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% DEFORMATION We have seen results of honeycomb testing on other plant dockets which indicate that the force-defomation curve becomes non-linear and increases rapidly at approximately l
40-50% defomation.
Percent defomation is defined as change l
in height divided by the original height.
Point P above H
l indicates the end of the horizontal portion of the curve.
You have indicated in FSAR section 2.6.2.2 that you use a deformation limit of 50% for honeycomb.
If, for example, the honeycomb used at Midland exhibited the above increase in the force-deformation curve after 40%
defomation (P = 40%), then your 50% deformation limit would y
allow the honeycomb to be crushed into the rapidly increasing portion of the force-defomation curve.
In such a case the honeycomb would be much stiffer than your analyses had assumed.
Consequently, the honeycomb would transmit a higher load to the supporting structures and concrete than your analysis would predict.
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-y 110-5 110.45 The staff feels that a prudent method for avoiding this potential problem is to set your deformation limits such that your honeycomb will only be crushed in the horizontal portion of the force-deformation curve, i.e., not past P
- H The force-deformation diagram provided in FSAR section 3.6.2.2 implies that the honeycombs used at Midland will indeed exhibit horizontal force-deformation characteristics up to and past your 50% deformation limit. Therefore, in order for the staff to complete its review, you should:
l.
Review the specific force-deformation curves for the honeycomb restraints used at Midland and verify that they all exhibit horizontal force-deformation characteristics up to and past the 50% deformation limit.
2.
If not, revise your deformation limits to assure that your honeycomb will only be crushed in the horizontal portion of the force-defonnation curve.
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121-1 121.0 MATERIALS ENGINEERING BRANCH - MATERIALS INTEGRITY SECTION 121.17 Compare all tests, data, methods, proposed programs, etc.,
(3.2) as presented in the Midland Unit Nos. 1 and 2 FSAR, Technical Specifications, and any other referenced sources (such as topical reports) on a point by point basis with the require-ments of Appendix G, " Fracture Toughness Requirements,"
and Appendix H, " Reactor Vessel Materials Surveillance Program Requirements," of 10 CFR Part 50.
Identify all areas of non-compliance to the Appendices.
121.18 It is our position that all of the procedures recommended (App. 3A)
(RSP) in Revision 1 to Regulatory Guide 1.99 be used in evaluating radiation damage to the reactor vessel materials of Midland Unit Nos. 1 and 2, unless an acceptable alternative method, with complete technical justification, is provided.
State your intent with regard to this position.
121.19 We have reviewed the information presented in FSAR (5.3.1.6)
(RSP)
Section 5.3.1.6, and we have concluded that the proposed change in the materials surveillance program is acceptable.
Therefore, it is our position that weld metal WF-209 may be substituted for weld metal WF-70 in the reactor vessel materials surveilllance program for Midland Plant Unit No. 1.
You should state in the FSAR that the available WF-70 weld metal will be stored in the material archives.
If you
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~N 121-2 should propose to use this weld metal in a future testing program of in a revised materials surveillance program, you should submit the details of the program for our evaluation and approval.
i This staff position will entail a formal exemption to Commission regulations.
This exemption will be issued after the response to request 121.17 has been evaluated.
This evaluation and the issuance of exemptions must be complete before the Midland Plant Unit 1 can be granted an operating license.
I 121.20 Babcock and Wilcox has informed us (IE Preliminary (5.3.1)
(RSP)
Notification of Event of Unusual Occurrence PNO-78-141A) l that certain B4W supplied reactor vessels may have been fabricated with weld wire that was not within specification.
This weld wire is characterized by an atypical composition l
having high silicon and low nickel.
l B6W has stated that a portion of weld filler wire heat number 72105 was mis-identified by the weld wire manufacturer.
This weld wire was used to weld beltline materials, ABZ-196 l
and ACA-197 (weld qualification number WF-70), in the Midland l
l Plant Unit No. 1 reactor vessel.
B6W has determined that the ductile to brittle tr7nsition temperature of the faulty weld wire may be 52 co 100 F higher than that measured on a properly fabricated WF-70 weld.
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121-3 It is our position that Consumers Power Company verify the presence or absence of this weld wire in the Midland Plant Unit No. I reactor vessel.
If it is found that this weld wire was used to fabricate the reactor vessel, then you must re-evaluate the technical specification pressure-temperature limit curves to show conformance with the requirements of 10 CFR Part 50, Appendix G, " Fracture Toughness Requirements."
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I 231-1 231.0 CORE PERFORMANCE BRANCH - FUEL DESIGN SECTION 231.24 Regarding your response to request 231.2 concerning stress and (4. 2.1.1) strain limits:
a.
It is not yet clear now the 0.4% elastic strain limit is derived and how it is related to Condition I and II (or III) events.
For example, how is the 0.4% elastic strain limit met in a Condition II event such as " Control Rod Withdrawal"?
b.
The stress and strain limits provided in FSAR Section 4.2.1.1.2 appear to be intended for the fuel system in general.
What stress limits are proposed for individual components of the fuel system such as spacer grids, guide tubes, fuel rod cladding, control rods, and other fuel system structural members of Condition I, II, III, and IV events?
231.25 Additional information is needed associated with our (4. 2. 3.1) request 231.3 and our review of FSAR Section 4.2.3.1:
a.
Please show, by means of an exemplar calculation, how combinations of system operating transients are evaluated to ensure that the " cumulative usage factor is less than 0.9 of the allowable material fatigue life."
Also show, by use of examples (as originally requested in request 231.3), under what circumstances the O'Donnell and Larger fati;ue curve is reduced by a factor
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231-2 of two on stress, rather than 20 on the number of cycles, to derive the design fatigue curve.
b.
The discussion of fretting in FSAR subsection 4.2.3.1.4 alludes to some testing performed in the B6W control rod drive facility and also to some post irradiation examination of fuel assemblies.
In neither case is a report number or other reference provided in support of the rather brief FSAR discussion.
Please provide the desired references.
c.
Recent excessive wear problems have been encountered i
in some B4W plants equipped with Mark B fuel assemblies with burnable poison rod assemblies and orifice rod assembly " ball-lock" coupling devices.
Please discuss what remedial design changes are being made to preclude similar problems in Midland Units 1 5 2.
Please state the allowable fretting wear and indicate the relationship to the cumulative usage factor.
231.26 The response to request 231.4 concerning (a) the specification (4.2.1.1) for dryness of the fuel pellets, (b) the statistical sampling technique, and (c) the method of moisture detection is inadequate.
A numerical value for the moisture limit and its relationship to the operating experience referred to in the response to request 231.4 (FSAR Section 4.2.1.1.4) is needed.
Provide typical values for the number of c
/
b 231-3 pellets per rod analyzed.
Also discuss the method used for moisture detection.
231.27 The'last sentence of FSAR Section 4.2.1.1.1 " Mechanical (4.2.1.1)
Properties" implies that Zircaloy cladding property values provided in Chapter 15 for transient analysis are not the same as those provided in FSAR Chapter 4.
If this is correct the diiferences should be identified and the rationale for such differences should be provided.
Also provide a tabulation of the UO and Zircaloy properties, along 2
with a list of data sources supporting these values (as requested in request 231.5).
231.28 Please provide the chlorine and fluorine limits used in (4.2.1.3) fuel rod fabrication and compare these to the inpurity levels known to affect cladding performance.
231.29 The stress on the holddown spring (102,800 psi) provided (4.2) in Table 4.2.7 of FSAR Rev. 9 is close to the stress intensity limit (107,000 psi).
Please show how these values were calculated and discuss why you conclude the rather small margin is adequate for your design.
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231.30 The response to request 231.20 concerning pellet / cladding (4.2.3.1) mechanical interaction, implies that cladding mechanical interaction failures are simply a function of* residual
-ductility (see FSAR subsection 4.2.3.3.1).
If the cladding
q 231-4 were totally free of defects, then this assertion would be valid.
However, after the fuel has experienced significant burnup, it will have " defects."
Moreover, unless the same stress and environmental conditions exist during a laboratory test as exist in a fuel element in the reactor, the test results are not directly applicable, but require interpretation.
Therefore, your FSAR statement that "recent irradiated cladding ductility data... indicate that current production cladding operating in the temperature range typical of a PWR (550F or greater) retains a ductility during irradiation in excess of that which would lead to a PCI concern" is misleading. Please modify or eliminate this statement.
Additionally, request 231.20 was not answered satisfactorily in regards to our request for identification of the operating conditions that would lead to a PCI failure concern (even if the referenced test data were accepted as directly applicable).
We are studying pellet / cladding interaction generically.
There is evidence that PCI is a concern not only during normal operation but for transients and accidents as well.
Your discussion of pellet / cladding interaction (PCI) of the kind initiated by stress corrosion cracking, provided in FSAR subsection 4.2.3.1.3, addresses normal operation experience only:
No discussion is provided
n 231-5 concerning the potential for PCI under transient and accident conditions, including ATWS.
Particularly for transient overpower or reactivity insertion events, where the fuel pellets may overheat and expand against the cladding, the potential exists for fuel failure due to PCI.
Because our review of the consequences of PCI failures has so far not resulted in the identification of specific safety problems, we have not imposed any operating restrictions.
If any safety issues are identified in the future, however, appropriate restrictions will be implemented.
231.31 The response to request 231.21 in FSAR subsection 4.2.3.4.2
[4.2.3.4) is very general and lacks detail.
For example, although a brief reference is made to analyses of frictional forces between fuel rods and spacer grids, no detail is provided concerning the magnitude of these forces (as compared to hydraulic forces) or the limits to irradiation growth of fuel rods and spacer grids.
Nor is any detail provided regarding the post-irradiation examinations that are asserted to support the conclusion that the frictional forces are adequate throughout life.
Please provide sufficient detail to support your response.
231.32 The response to request 231.23 in FSAR subsection 4.2.3.5.3 (4.2.3.5) should include some numerical values for measured fuel handling and shipping loads for comparison with design loads.
Please provide the measured and design values
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o 231-6 so that we can assess the margins available.
231.33 Your FSAR contains a fairly detailed description of the (4.2.1.3) models in the TACO fuel performance code, but is not clear as to what application is made with TACO.
Clarify your FSAR to indicate that TACO is used for all safety analyses (including ECCS analyses) related to reactor licensing; also, verify that previously used codes like TAFY have not been used on Midland.
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232-1 232.0 CORE PERFORMANCE BRANCH - PHYSICS SECTION 232.12 Your response to request 232.2 is inadequate.
Topical (4.3.2) report BAW-10028 indicates that certain misloadings in equilibrium cycles could lead to violation of fuel limits or DNBR.
Does this conclusion apply to the Midland first cycles?
Ifso,couldsuchmisloadingsbedetectekbythe incore instrumentation?
Provide a quantitative discussion including the effect of the misloading on the misloaded assembly and on neighboring assemblies containing incore detectors.
Provide conclusions regarding the probability of detection of assembly loading errors and, if appropriate, discuss the consequences of operating with ar andetected misloading.
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331-1 331.0 RADIOLOGICAL ASSESSMENT BRANCH 331.6 Revision 11 of the FSAR deleted the requirement for (12.5.2.1.1) having two portal monitors and a frisking station at the exit point of controlled areas (as mentioned in the second sentence of FSAR Section 12.5.2.1.1).
Describe what degree of monitoring you will require at the exit to controlled Zones.
331.7 Revision 11 to Section 12.5.3.3 of your FSAR states that (12.5.3.3) radiation workers will be tested for their understanding of radiation protection principles once every two years, rather than once every year as was stated in the ySAR prior to Revision 11.
Provide your justification for this two-year testing interval.
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362-1 362.0 GEOTECHNICAL ENGINEERING 362.9 The response to request 362.4 is insufficient.
Table 2.5-14A (2.5.4) shows the structur.al settlement measurements available to date.
Provide the reasons for the lack of survey data at benchmark numbers A-3 and 4, C42, 3, 4, 5, 6 and 7, and T-2, 5, 6, 7, 8, 9, 10, 11, 12, 13, 14 and 15.
In section 2.5.4.13.1 of the FSAR, reference is made to Figure 2.5-78.
The figure number is in error and should be corrected.
362.10 The SER on the PSAR stated that continued surveillance for (2.5.4) subsidence sheuld be maintained throughout the life of the plant.
Provide in Section 2.5.4.13 of the FSAR a discussion on the scope and details of the subsidence monitoring program.
Include a commitment to monitor subsidence throughout the life of the plant, and indicate the proposed survey frequency.
Submit all subsidence data measured since installation of the benchmarks.
c 400-1 400.0 PROJECT RNAGEMENT
'400.3 The following regulatory guides reflect current NRC (3A) staff practice.
Except in those cases in which you propose an acceptable alternative method for complying with specified portions of the Commission's regulations, the methods described in these guides are being and will continue to be used in evaluation of submittals in connection with your application for operatig licenses.
This includes original issues which are stamped "for comment."
Discuss conformances to these regulatory guides in Appendix 3A of your FSAR:
RG DATE i
1.109 October 77 1.111 July 77 1.112 May 77 1.113 April 77 1.115 July 77 1.116 May 77 1.122 September 76 1.123 July 77 1.126 March 77 1.127 March 78 1.133 September 77 1.134 September 77 1.140 March 78 1.141 April 78 1.143 July 73
422-1 422.0 CONDUCT OF OPERATIONS 422.8 Your response to our position stated in part a to (13.1.3.1)
(RSP) question 422.5 is not satisfactory.
It is our position that the individual to whom the Plant Supervisors report, a position comparable to that described in subsection 4.2.2 of ANSI N18.1, should, at the time of initial core loading, hold a Senior Reactor Operators' License.
Revise your response accordingly.
422.9 Your response to our position stated in part b to question (13.1.3.1)
(RSP) 422.5 and the information submitted in regard to your conformance to Regulatory Guides (specifically, Regulatory Guide 1.8) in Appendix 3A of your FSAR does not clarify your intended qualification requirements for the position of Health Physicist.
It is our position that an individual on your plant staff meet the qualifications for the position of Radiation Protection Manager described in Revision 1 to Regulatory Guide 1.8, September 1975, i.e.,
should have a Bachelor's Degree or the equivalent in a science or engineering subject, including some formal training in radiation protection.
He or she should have at least five years of professional experience in applied radiation protection.
(A Master's Degree may be considered equivalent to one year of professional experience, and a Doctor's Degree may be considered equivalent to two years of professional experience where course work related to radiation protection is involved.)
At least three years of this professional
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422-2 in a nuclear facility dealing with radiological problems similar to those encountered in nuclear power stations, preferably in an actual nuclear power stat'on.
Revise i
your response accordingly.
1 422.10 The qualification requirements you describe for the (13.1.3.1)
(RSP) position of Quality Control Supervisor are not satis-factory.
It is our position that the qualification requirements for this pcsition should be six years experience in the field of quality. assurance, preferably at an operating nuclear plant or operations supervisory experience.
At least one year of this six years experience shall be nuclear power plant experience in the'overall implementation of the quality assurance program.
(This experience shall l
be obtained within the quality assurance organization.)
A minimum of one year of this six years experience shall be related technical or academic training.
A maximum of four years of this six years experience may be fulfilled by related technical or academic training.
Revise your FSAR accordingly.
422.11 The quorum requirements for your Plant Review Committr a (13.4.2.1)
(RSP)
(PRC' are not satisfactory.
It is our position that the quorum requirements should include the Chairman or Alternate Chairman plus five members.
Revise your FSAR accordingly.
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422-3
' 422.12 In regard to the' subjects to be reviewed by your Safety (13.4.3.8)
(RSP)
Review and Audit Board, they should include "all recognized indications of an unanticipated deficiency in some aspect of design or operation of safety-related structures, systems, or components," in accordance with the Standard Technical Specifications.
ReviseyourFSARaccorkincly.
422.13 You describe in subsection 13.4.3.9 your provisions for (13.4.3.9)
(RSP) conducting technical audits.
It is our position that the audit program should include the following items:
a.
The results of actions taken to correct deficiencies occuring in facility equipment, structures, systems, or method of operation that affect nuclear safety at least once per 6 months.
b.
The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix B, 10 CFR 50 at least once per 24 months.
c.
The Facility Security Plan and implementing procedures at least once per 24 months.
d.
The Facility Fire Protection Program and implementing procedures at least once per 24 months.
e.
An independent fire protection and loss prevention inspection and audit shallibe performed annually
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422-4
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l utilizing either qualified offsite licensee personnel or an outside fire protection firm.
f.
An inspection and audit of the fire protection and loss preventaion program shall be performed by an outside l,
qualified fire consultant at intervals no greater than
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3 years.
g.
Fire Protection Program implementation.
Revise your FSAR accordingly.
422.14 Describe the composition (including. number of per' sons)
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(13.1.2) of your plant fire brigade.
This should include that period when only the first of the operating units is operating, 1
and that period when both units are operating.
Our position for such manpower requirements is given in Mr.
D. Vassa11o's letter of August 11, 1978 to Mr. S. H. Howell.
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441-1 441.0 OPERATOR LICENSING BRANCH - TRAINING SECTION 441.8 Amend FSAR Section 13.2 to include fire protection training (13.2) as required by revision 1 of Standard Review Plan 13.2.
The applicable sections of SRP 13.2 that pertain to fire protection training are I.5, II.2, II.6, III and IV.
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372-1 372.0 METEOROLOGY 1
3'72.18 The response to request 372.01a is inadequate.
(2.3.1)
According to the 1971 study by M.A. 3111ello entitled, " Frozen Precipitation:
Its Frequency and Associated Temperatures", the mean monthly density of snowpack at Oscoda, Michigan was estimated to be 0.3 g/cm3 In aramination of Table I, Figure 2, Figure 3 and Figure 6 of the Bilello reference, the value of 0.3 g/cm3 appears to be applicable to the Midland site. However, for calculations of the weight of snowpack at the Midland site, a snowpack density of 0.25 g/cm3 was assumed. Justify why the value of 0.25 g/cm3 was used instead of a larger value.
372.19 The atmospheric dispersion model and procedures used to evaluate (2.3.4) dispersion conditions to be used in an assessment of the consequences of design basis accidents described in Section 2.3.4 are based on Regulatory Guide 1.4 and Section 2.3.4 of the Standard Review Plan.
Af ter review of the results of recent atmospheric dispersion field experiments, we have developed a modified procedure for calculating short-term relative concentration (X/Q) values wnich considers the following:
(1) lateral plume meander; (2) atmospheric dispersion conditions as a function of direction; (3) wind direction frequencies; and (4) exclusion area boundary distances as a function of direction.
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372-2 Enclosed is a copy of DRAFT Regulatory Guide 1.III, " Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants" (9/23/77), which describes the new procedure in detail. We believe that this modal vill provide an improved characterization of atmospheric dispersion conditions around the Midland site. Also enclosed is the interim branch technie=1 position 4
concerning use of these two models. During our review, we will
- wamine I/Q values for appropriate time periods for design basis accident evaluations using the modified model described in the enclosed DRAFI Regulatory Guide, and compare them with I/Q values enleninted using the model described in Regulatory Guide 1.4 and the procedures described in Section 2.3.4 of the Standard Review Plan. Therefore, provide avelusion area boundary distances as a function of direction using the protiedure described in the DEAFI Regulatory Guide.
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ATTACHMENT TO REQUEST 372.19
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HMB-BTP-2 (REVISION I) f (AUGUST 2,1978)
INTERIM BRANCH TECHNICAL POSITION HYDROLOGY-METEOROLOGY BRANCH ACCIDENT METEOROLOGY MODEL It is our position that either the draft Degulatory Guide 1.XXX,
" Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants" (dated September 23, 1977),
or the procedures described in Standard Review Plan Section 2.3.4 may be used to evaluate atmospheric transport conditions for analysis of accidents with the following amendments to the draft regulatory guide model:
(a) a limiting sector X/Q value at the 0.5% probability level be used*,
(b) the accumulated frequency of the limiting sector l
X/Q or higher valise in all sectors may not exceed 5% for the site, and; (c) norma 7'.zation of individual sector probability distributions is not used.
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- Amendment based on Memorandum from H. R. Denton to D. R. Muller,
Subject:
Proposed New Meteorological Model, dated August 2,1978.
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