ML19326D413
| ML19326D413 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 02/24/1978 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | Howell S CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| NUDOCS 8006110429 | |
| Download: ML19326D413 (80) | |
Text
~
[....
UNITED S TATES f
g 7,,
NUCLEAR REGULATORY COMMISSION f -}. ),JJ).. l
-i W ASHINGTON, C. C. 20555 UNl1%fl a
p February 24, 1978 Cocket Nos. : 50-329 & 50-330 Consumers Pcwer Company ATTN: Mr. S. H. Howell vice Fresident THIS DOCUMENT CONTAINS 212 West Michigan Avenue P00R QUAL.lTY PAGES Jackson, Michigan 49201 Gentlemen:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION - PART ONE Ir continuing our review of the FSAR for Midland Plant Units 1 & 2, we find we need additional information to complete our evaluation.
This infor-mation request is contained in Enclosure 1.
Tha information requests provided in Enclosure 1 use a sequential nudering system continuing from those folicwing our acceptance review and provided by our letter of Novembar 11, 1977. As indicated in our letter of December 27, 1977, we have scheduled our round-one requests in three sepa ate parts for which this is the first part. Enclosure 1 is based upon our revien of FSAR revision nutbers three or fcur.
We will need complete and adequate responses te Enclosure 1 by April 14, 1978.
If you cannot meet this date, inform us witnin seven days after, recaipt of this letter so that we may revise our schedule accordingly.
Some of our requests also represent Regulatery Staff Positions and are identified by the initials RSP.
If, during the course of our review, you believe there is a need to appeal a staff position because of dis-agreement, this need should be brought to our attention as early as possible so that the appropriate meeting can be arranged on a timely basis.
A written request is not necessary and all such requests should be initiated through our staff project manager assigned to the review of your application.
This procedure is an informal one, designed to allow oportunity for applicants to discuss, with management, areas of disagreement in the case review.
Please contact us if you desire clarification or other discussions of the information requested.
Sincerely, 1 &
s S. A. y rga, Chief 2
Light Water Reactors Branen No. 4 Divisio6 of Project Management
Enclosure:
As Stated 80 0 61 y,0 8
a.
~
l 3
. ~. -.
.l a
A
_ r.
j Consurners Pcwer Certpany ces:
Micha41. I. Miller, Esq.
Lee Mute, Esq.
-Isham, Lincoln & Beale Michigan Division Suite 4200 The Ocw Chemical Company One First National Plaza 47 Building Cnicago, Illinois 60670 Midland, Michigan 43640 Jucd L. 3 acon, Esq.
Managing Attorney Consumers Power Company 212 West Michigan Avenue Jackson, Micnigan 49201 Mr. Paul _A. Perry Secretary Censumers Power Coinpany 212 W. Michigan Avenue Jackson, Michigan 49201 Howard J. Vogel, Esq.
Knittle & Vogel 814 Flcur Exchange _auilding Minneapolis, Minnesota 55415 Myron M. Cherry, Esq.
One IBM Plaza
. Chicago, Il I inois 60611 Honorable Curt -Schneider Attorney General State of Kansas Topeka, Kansas 66612 Irving Like, Esq.
Reilly, Like and Schneider 200 West Main Streat Babylon, tiew York ~ 11702 James A. Kende11, Esq.
.Curtie and Kendall 135 North Saginaw Road
-Midland, Michigan 43640 Louis W. Pribila, Esq.
. Michigan Divisicn Legal :>epart. ment
' 47 Building Dow Chemical USA Midland, Michigan 48640 l
IL
~. _.
_j
.n e
,s REQUEST'FOR ADDITIONAL INFORMATION (Qls)
PART 1 of 3 MIOLAND PLANT UNITS 1 & 2 These requests for additional informacion are numbered such that the
~
three digits to the left of the decimal identify the technical review branch and the numbers to the right of the decimal are the sequential request numbers. The number in parenthesis indicates the relevant section.in the Safety Analysis Report. The initials RSP indicate the request represents a regulatory staff position.
Branch Technical Positions referenced in these requests can be found in " Standard Review f ian for the Review of Safety Analysis Reports for Nuclear Power Plants," NUREG-75/087 dated September 1975.
e
.am Y
f' p
y-010-1 010.0 AUXILIARY SYSTEMS BRANCH 010.19 Your response to 'our request '010.2 is not cceplete. Provide the
-(3.2.4) following information:
.a.
Provide the bases for assuming that a flood from circulating water system failure will be limited to the plant grade level at elevation 20'-6" inside and outside the turbine building areas. Describe the turbine building wall construction and door arrangement to demons,trate that flood water inside the.
turbine building will not incapacitate safety related equipment.
b.
List all doors and other openings on the walls between turbine building and auxiliary building below the grade level.
Describe the water-tight design of these doors and openings and their related administrative controls to ensure that they are closed when isolation is needed.
010.20 Provide layout drawings of the safety-related arear outside contain-(3.6) ment showing all high ar.a moderate energy piping systems and their relation to the safety related equipment.
010.21 Your response to our request 010.6 is not complete. Describe the (9.0) automatic heating system for the borated water storage tank in sufficient --
detalk including equipment and instruments design classification and power sources. Also, confirm that the borated water heating system design meets the single failure criteria and is Class lE.
~010.22 Your response to our request 010.10 is not acceptable. Provide I9 I 4) detailed artan9ement.dF.awings of the fuel pool and fuel cask handling areas including areas housing safety-related equipment below the cask handling path. Modify your design to meet the following
)
positions:
j
~
m, 010-2 010.22 a.
Provide both mechanical and electrical stops for the cask handling crane to preclude possible passing of heavy loads over the fuel pool.
b.
Provide the results of an analysis to demonstrate that the floor above safety related equipment can withstand a cask drop from the maxinum possible cask lift height assuming a cask drop configuration that will result in the worst effect to the floor and not perforate the floor or generate secondary misailes on the other side of the floor that would damage safety related equipment.
Provide physical restrictions, including mechanical or electrical c.
interlocks, in addition to the administrative controls to ensure that the cask lifts will not exceed the maximum elevation assumed in your cask drop analysis.
010.23 Section 9.2.1 and Table 3.2-1 seem to indicate that the service 4
(9.2.1) water traveling screens and screen wash pumps are not designed to seismic Category I requirements and not connected with essential power supplies. Provide the results of an analysis to-demonstrate that during post-accident or following a. tornado or seismic event with loss of offsite power, the essential service water intake
~
will not be blocked due to accumulation of debris or modify your design accordingly.
010.24 Section 9.2.1 indciates that the service water pump pit normally (9.2.1) receives water supply from the cooling towers. An engineered safety features actuation signal (ESFAS) will automatically shift the water supply to the cooling pond by opening the power operated valves at the discharge lines to the pond, and opening the power operated sluice gates-between the pump pit and the pump house forebay.
Storage capacity of the pump pit is adequate to feed all four service water pumps while the sluice gate is being opened.
~#~"
- a r
e-
=
g 010-3 1...
010.24 Provide.the results of an analysis to demostrate that the service water pumps arc protected from damage due to low suction pressure when the non-safety grade water supply from cooling towers is lost (e.g., SSE) without presence of the ESFAS.
If operator
. action is required in this case, 30 minutes time should be assumed.
Confirm Class lE service water pump pit level indication and alarms are available inside the control room.
010.25' Figure 9.2-1 and Figure 3.5-1 indicate that the essential service (9.2.1) water supoly and return lines are tornado missile protected by being buried or located in seismic Category I buildings. Provide results of an analysis to demonstrate that the depth of the buried piping is sufficient to protect the safety related service water piping from tornado generated missiles in the area.
010.26 Your response to our request 010.12 is not acceptable.
It is our I9 j position that you must modify the design of the component cooling water system supplying cooling water to tne reactor coolant pumps to meet the position stated in our request 010.12.
If you plan to demonstrate that the RCP's of fiidland Plant, Unit pos.1 & 2 will not excerience shaft seizure following loss of coolino water for longer than 30 minutes without the need for operators corrective action, then an actual pump test should be conducted for verifica-tion. Also, 'to satisfy the criteria in Approach 1 of the above position, safety grade instrumentation to detect the loss of CCW to the RCP's and to alarm to the operator in the control room should be provided.
010.27' Modify your Figure 9.2-19 to show the auxiliary feedwater supply (9.2.6) lines from the condensate storage tanks.
4 010.28 Expand Section _9.2.6 to _ discuss the basis. for ' sizing the condensate (9.2.6)'
storage tanks including the minimum condensate storage available for auxiliary feedwater supply. State assumed time period for plant hot standby and cooldown using condensate supply. Describe v
=
w g
g u
- - - - +
Yr
e 010-4 010.28 how the minimum condensate storage capacity is maintained in each condensate storage tank for the auxiliary feedwater system.
~
010.29 Design deficiences have been identified in other B&W plants such that I9'3*4) following a loss of offsite power,* the reactor coolant pump seals cannot withstand the resulting interruption of seal water flow without damage. Expand Section 9.3.4 to address the Midland Plant design relative to the above stated deficiencies. Describe the modifications made to correct the deficiency if there is any, and confirm that the Midland plant can withstand a loss of offsite power without seal damage to the Reactor Coolant Pumps.
010.30 Section 9.4.1 states that the battery room HVAC system is designed U
to maintain the hydrogen concentration below 4.1 volume percent.
This is not acceptable.
It is our position that you redesign the battery room ventilation system to limit the hydrogen concentration to well below two volume percent and clarm in the control room when battery room ventilation is lost.
010.31 Figure 9.4-3 indicates that there are a 22-inch and a 25-inch non-(9.4.2) seismic Category I air supply ducts passing through the spent fuel pool area without se.ismic Category I isolation danpers in the lines.
Provide the results of an analysis to demonstrate that the operation of the standby exhaust and filtration system af ter a postulated fuel accident will maintain its design safety function assuming the spent fuel pool structure boundary is broken due to non-seismic Category I duct failure.
010.32 Section 9,4.3 and Section 9.4.5 indicate that the HVAC system for the I
Auxiliary area is not designed to seismic Category I requirements and yet safety grade unit coolers are provided in each ESF equip-ment rocm for control of environment. Provide a discussion
p 010-5 to demonstrate how, without a safety grade exhaust and filtration system in the ESF equipment area, a necative cressure can be maintained inside these equipment rooms to preclude possible radio-active release into the environment during post-LOCA operation, or provide a seismic Class I exhaust and filtering system.
010.33 Your response to our request 010.18 is not complete. Provide the (10.4.9) following ' additional infomation:
a.
Section 10.4.9.2 indicates that the AC operated backup cool-ing water supply to the turbine driven auxiliary feedwater pump is required when the temperature exceeds 100*F. Veri fy and confim that the temperature will not exceed 110*F.
during emergency operation using condensate or service water supply at the maximum pcssible water temperature during summer season.
b.
Figure 10.3-1 and Figure 10.4-10 indicate that the AC power operated valves in the turbine driven auxiliary feedwater sub-system may f ail in the cpen position.
Discuss the failure mode of these AC valves and confim that the turcirie driven auxiliary feedwater subsystem can operate without AC power supply _to meet the diversity requirements of our Branch Technical Position APCSB 10-1.
c.
Section 10.4.9.3 states that the reactor coolant temperature can be reduced to about 310*F when using the turbine driven auxiliary feedwater-pump (TDAFP). At this point, the TDAFP is stopped and the steam generators boiled down to reduce the reactor coolant temperature to 230*F, at which coint the decay heat removal system can operate. This reductior, from 310 to 280*F can be accomplished by dumping heat to the main con-denser and circulating water system. Assuming a failure of the motor-driven auxiliary feedwater pump and that the main condenser is lost.(e.g., loss of offsite power), confirm tnat safe cooldown of the plant can still be achieved.
b
e t ~
[
p.
010-6 010.34 Your response to our request 010.14 is unacceptable.
You depend I
5P) n manual remote control of valves from the service water system to
- Exiliary feedwater pump suction following a loss of condensate storage tank supply. There are two events which would lead to unacceptable consequences as a result of your design, since the condensate storage tank is not seismic Category I or missile protected. These are:
a.
A seismic event could result in the failure of the con- '
densate storage tank and the loss of offsite power.
The auxiliary feedwater pumps would start with'no NPSH resulting in the loss of auxiliary feedwater capability; and b.
A tornado could result in a steam line break in the unpro-tected non-safety portion of the steam line, loss of offsite pcwer, and the loss of the condensate storage tank. Again, the auxiliary feedwater pumps would start with no NPSH resulting in the loss of the auxiliary feedwater system.
~.-
It is our position that you protect the plant against these events.
Provide a seismic Categoty 1, tornado protected condensate storage tank, or provide an automatic switchover to the service water syster using safety grade instrumentation and demonstrate that sufficient auxiliary feedwater flow will be available in the time required to
~
prevent unacceptable consequences following these events.
010.35 The design of your auxiliary feedwater system consists of one 00.4.9) motor-driven pump (100".) and one turbine driven pumo (100"). A high energy line break at the discharge of one pumo and a single active failure of the other pumo will result in the inability to perform a safe plant shutdown and cooldown. Revise your design to withstand a high energy line break, coincident with a single active failure in the auxiliary feedwater system.
~
g3 n
022-1 022.0 CONTAINMENT SYSTEMS BRANCH 022.6 You state in Section 6. 2.1.1. 3. 7 that the instrumentation provided (7.5, to monitor and record containment parameters following an accident 4
6.2.1.1, does not include the containment emergency sump water temperature.
6.2.2.1.5, We require that instrumentation be provided to monitor and record 6.3.5) this parameter. Revise your design and discuss your intended RSP compliance with this staff position.
022.7 Section 6.2.1.1.1.6.5, Passive Heat Sinks, states that.the heat sinks (6.2.1) used in the minimum containment pressure analysis for the ECCS evaluation are listed in Table 6.2-10.
However, the heat sinks in Table 6.2-10 are also used for calculating the maximum containment pressure. The heat sinks used for both types of analyses should reflect a degree of conservativism to account for the uncertainty in developing the data; i.e., for the minimum containment pressure analysis, conservatively high values should be used, and for the maximum containment pressure analysis, conservativiely low valves should be used. Therefore, discuss how the
. heat sink data was developed, and its applicability to the maximum and minimum containment pressure analyses.
If necessary, revise the analyses using appropriately conservative heat sink data.
'022.8.
Your analysis in Secticn 6.2.1.1.3.6 of inadvertent actuatieri of (6.2.1) the containment spray system to determine the containment external design pressure assumes a heat transfer coeffecient from the contain-2 ment shell to the containment atmosphere of 2 BTV/hr-f t 'F.
Justify this value.
022.9.
The Engineered Safety Features Actuation System (ESFAS) provides the (6.2.4, signals for containment isolation, which is only high containment 7.3) pressure. We require diversity in the parameters sensed for the initiation of containment isolation. Therefore, discuss your plans for including other ESFAS; e.g., signals to provide the required diversity for containment isolation.
022.10 In Table 6.2-3 and Section 6.2.6, specify the maximum allowable (6.2.6) containment leakage rate (La) in weight percent per day.
022.11-Identify the containment isolation arrangements which do not comply
.(6.2.4) with.the explicit. requirements of General Design Criteria 55, 56 and 57, and discuss the rationale for concluding that the isolation arrangements
-are acceptable on some other defined basis.
m.4
n
.m
~
022-2 022.12 Provide the following information regarding tho hydrogen. production and (6.2.5.3) accumulation analysis:
- a. Discuss.the applicability of. the experimental data used to support the corrosion rates selected for the aluminum, galvanized materials, and zine base paints listed in Table 6.2-3a.
Idcatify the key parameters which influence the corrosion rates. Compare the parameters that would be expected in the Midland containment folicwing a LOCA to the parameters 'or the experimental data.
- b. Discuss the converatism in the c.antities and surface areas of the galvani:ed steel and zinc base paint assumed in the analysis.
Also, discuss the rationale for not considering aluminum base paint in the analysis.
022.13 Identify those fluid lines penetrating the containment which will be (6.2.6.1) vented and irained to ensure exposure of the system containment isolation valves to tae containment atmosphere and the full differential pressure during the containment integrated leakage rate (Type A) test. Those systems that will remain fluid filled for the Type A test should be identified and justified.
022.14 For each fluid line that penetrates the containment, schematically (6.2.6.3) show the isolation valve arrangement and the design provisions (e.g.,
test, vent and drain connections, black valves) that will permit the isolation valves to be leak tested Indicate the direction in which the valves will be leak tested.
Identify, in Table 6.2-23, all valves for which the applied test pressure will not be in the same direction as the pressure existing when the valve is required to perform its safet function, and provide evidence to show acceptability of testing the valve with pressure applied in the reverse direction.
~
022.16 10 CFR 50 Appendix J requires that containment penetrations fitted with (6.2.6.3) expansion bellows be locally tested at the calculated peak containment pressure, Pa.
Identify the penetrations fitted with expansion bellows and verify that this requirement can be met.
.022.16 Table 6.2-28 identifies the containment isolation valves that will (6.2.6.3) not be subject to Type C leak testing; for example, locked-closed containment isolation valves. ~ Also most of the containment isolation valves under General Design Criterion 57 (closed systems inside contain-ment) will not be subject to Type C leak testing. Discuss your plans to Type C test these valves or justify exempting them frem Type C testing.
O
~
022-3 022.17 Your respense to request 022.2F.did not dicusss the nadalization sensitivity:
(6.2) study performed for each subcompartment to determine the minimum nurber of volume nodes required to conservatively predict the loads acting on compartment walls and component supports. Provide this information.
Identify the'nodaliz& tion scheme used to calculate the loads acting on the compartment walls and that used to calculate loads on the components.
022.18 Provide the information requested in 022.2 fcr the reactor cavity
.(6.2) and pressurizer compartment for postulated ruptures in the following piping:
- a. Core flood tank lines.
- b. Cold leg piping in the reactor cavity.
- c. Pressurizer surge line,
- d. Pressurizer spray line.
.022.19 Our request 022.2 asked for analyses of the subcompartment pressure transien (6.2.1.2) used in the design of the component supports. Your response references RSP your letter of October 6,1977, which discusses your participation in a B&W Users Group to evaluate the probability of a reactor coolant system pipe rupture in the reactor cavity relative to our concerns regarding reactor vessel supports. Your letter also states that the results of1the probability study were submitted on September 27, 1977 by Science Applications, Inc. as Report No. SAI-050-77-PA. We find the approach described in the topical recort unacceptable and require the
' detailed analyses requested in 022.2. Provide the information requested in 022.2 for the reactor cavity, and other compartments subject to pressurization.
022.20 Your response-to request 022.5 is unacceptable. Discuss in detail how the
.(6.2) containment purge system design complies with the recom.mendations of our Branch Technical Position CSB 6-4 Also provide the analyses identified in Branch Technical Position CSB 6-4.
I e
"fffW
~
A 040-1 040.0 POWER SYSTEftS BRANCH 40.13 Identify all safety related cables used in your plants that have (8.3) polyethelene in its construction. Provide the following information for each type of cable using polyethelene:
Type of cable by name and catalogue number.
tianufacturer.
Type of polyethelene used and how, i.e., insulation and/or 1
' jacket.
Results of environmental qualification performed.
Identify the environmental qualification test report for each type of cable.
40.1a Recent operating experience has shown that adverse effects on the (8.2) safety-related power system and safety related equipment and loads RSP can be caused by sustained low or high grid voltage conditions.
We therefore require that your design of the safety related electrical system meet the four staff positions in attached Apptndix 40-1.
Supplement the description of your design in the FSAR
_r to show how it meets these positions or provide appropriate results of analyses to justify non-conformance with these positions.
I m
e em f
W
.. ~
n
.A 40-2 40.15 Recent repsrts of diesel generators at operating nuclear plants 8.3 reveal that in some cases the information available to the control room operator to indicate the operational status of the diesel generator may be imprecise and could lead to misinterpretation.
This can be caused by the sharing of a single annunciator station to alam conditions that render a diesel generator unable to respond to an automatic emergency start signal and to also alam abnomal, but not disabling, conditions. Another cause can be the use of wording of an annunciator window that does not specifically say that a diesel generator is inoperable (i.e., unable at the time to respond to an automatic emergency stait signal)' when in fact it is inoperable for that purpose.
Review and evaluate the alarm and control circuitry for the diesel generators for 'iidland Plant, Units 1 and 2 to deternine how aach condition that renders a diesel generator unable to respond to an automatic emergency start signal will be alamed in the control room. These conditions include not only the trips that lock out the diesel generator start and require manual reset, but also control switch or mode switch positions that block automatic start, loss of control voltage, insufficient starting air pressure or battery voltage, etc. This review should consider all aspects of possible diesel generator operational conditions, for example test conditions and operation from local control stations. One area of.particular concern is the unreset condition following a manual stop at the local station which teminates a diesel generator test and prior to reseting the diesel generator controls for enabling subsequent automatic operation.
Provide the details of your evaluation, the results and conclusions, and a tabulation of the following information-(a) All conditions that will render the diesel generator incapable of responding to an automatic emergency start signal for each operating mode as discussed above; (b) The wording on the annunciator window in the control room r
that will be alarmed for each of the conditions identified in (a);
(c) Any other signals not included in (a) above that also will cause the same annunciator to alam; (d) Any condition that will render the diesel generator incapable of responding to'an automatic emergency start signal which i
is not 31amed in the control rocm; and (e) Any proposed modifications planned as a result of your evaluation.
g
~
~
40-3 40.16 Describe how your electrical penetrations and associated connections (8.3) to the field cables. are qualified to withstand LOCA and Steam Line Break environment. Your response should address:
(1) Test Plan (2) Test set up (3) Test procedures (4) Acceptability goals and requirements.
Also, provide an evaluation of the results that demonstrate electrical oenetrations are qualified to rnaintain containment integrity during normal, abnomal, and accident conditions.
M' e
k i
i
~
+ + -
.yc M
y q
w
-9+----
--r
-w-v g-
c APPE:lDIX 40-1 STAFF POSITIONS ON VOLTAGE VARIATIONS Position 1: Additional Level of Under-or-Over Voltage Protection witn a Time Delay
'Je require that an additional level of voltage protection for the onsite power system be provided and that this additional level of voltage protection snall satisfy the following criteria:
a) The selection of voltage and time set points shall be determined from an analysis of the voltage requirements of the safety-related loads at all onsite. system distribution levels; b) The voltage protection shall include coincidence logic on a per cus basis to preclude spurious trips of the offsite power source;
.a.
c) The time delay selected shall be based on the following conditions:
(1) The allowable time delay, including margin, shall not ex eed the maximum time delay that is assumed in the PSAR accident analyses; (2) The time delay shall minimize the effect of short duration disturbances from reducing the availability of the offsite power source (s); and (3) The allowable time duration of a degraded voltage condition at all distribution system-levels shall not result in
~
failure of safety systems or components;
--m%%,..
O
b f
~
2 d) The voltage sensors shall automatically initiate the disconnec-tion of offsite power. sources whehenver the voltage set point and time delay limits have been exceeded; e)' The voltage sensors shall be designed to satisfy the applicable requirements of IEEE Std. 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations"; and f) The Technical Specifications shall include limiting condition for operation, surveillance requirements, trip set points with minimum and maximum limits, and allowable values for the second-level voltage protection sensors and associated time delay devices.
Position 2: Interaction of Onsite Power Sources with load Shed Feature We require that the current system designs automatically prevent load shedding of the emergency buses once the onsite sources are supplying power to all sequenced loads on the emergency buses. The design shall also. include the capability of the load shedding feature to be automatically reinstated if the onsite source supply breakers -
are tripped. The automatic bypass and reinstatement feature sha'.1
.tme verified during the periodic testing identif ed in Position 3.
3 a"
In the event-an adequate basis can be provided for retaining the load shed feature wnen loads are er.ergized by the onsite power system, we will require that the setpoint value in the Technical Specifications, which is currently specific as ".. equal to or greater than..." be amended to specify a value having maximum and minimum limits. The licensee' bases for the setpoints and limits selected must be documented.
Position 3: Onsite Power Source Testin; We require that the Technical Specifications include a test require-ment to demonstrate the full functional operability and independence of the ensite poner sources at least once per 18 months during
' " ~
shutdown. The Technical Specifications shall include a requirement for tests: (1) simulating loss of offsite power; (2) simulating loss of offsite pcwer in conjuncticn with a safety feature actuation signal; and (3) simulating interruption and subsequent reconnection of onsite power sources to their respective buses. Proper operation shall be determined by:
v g
~
4
,,rifying that on loss of offsite power the emergency buses have 7
- een de-energized and that the loads have been shed from the emergency buses in accordance with design requirements.
e) Verifying that en loss of offsite power the diesel generators start on the aut0 start signal, thc emergency buses are energized with permanently connected loads, the ;ato-connected shutdown loads are energized throug'.i the load sequencer, and the system operates for five minutes while the generators are loaded with the shutdown loads.
c) Verifying that on a safety features actuation signal (without loss of Offsite power) the diesel generators start on the autostart signal and operate on standby for five minutes.
d) Verifying that on loss of offsite power in conjunction with a safety features actuation signal the diesel generators start on the autostart signal, the emergency buses are energized with ;ermanently connected loads, the auto-connected emergency (accident) loads are energized through the load sequencer, and the system operates for five minutes while the generators are loaded with the emergency loads, e) Verifying that on interruption of the onsite sources the loads are
- shed from the emergency buses in accordance with design require-ments and that subsequent loading of the onsite sources is through the load sequencer.
i
..J '
'A e.
Position 4 - Ootimization of Transformers Tao' Settings The voltage levels at the safety-related buses should be optimized for the full load and minimum load conditions that are expected throughout the anticipated range of voltage variations of the offsite power source by appropriate adjustrent of the voltage tap
~
settings of the intervening transformers. We require that the adequacy of-the design in this regard be verified by actual measurement and by correlation of measured values with analysis results. Provide a description of the method for making this verification; before initial reactor power operation, provide the documentation required to establish that this verification has been accomplished.
. s.
ese u
h
R
~
O
~
110-1 110.0 MECHAtlICAL ErlGirlEERIrlG BRANCH
^
110.7 *.
Your response to request 110.1 is unacceptable. The staff will (3.6.2) require an augmented inspection program as outlined in request (RSP) 110.1. _ FSAR Section 3.6.2.1.4 should be modified accordingly.
110.8 Regarding your discussf'on of legitudinal break locations in FSAR (3.6.3)
Section 3.6 3:
(1) Ycur respon5c to request 110.2 is not entirely acceptable.
We are aware of 1.5e discussions on pipe break criteria as referenced in yout response. FSAR Section 3.6.3.1.b.2' discusses locationi at which longitudinal breaks need.wt ce postulated. We agree with this section with the excettien ofpartof(b). Epecifically, we require that both circum-ferential and longitudinal breaks be postulated at C1'ss 1 intemediate brea locations chosen because the cumulai. 3 usage factor exceeds 0.1.
(2) Similarly, we disagree with FSAR Section 3.6.3.1.b.1, part b, and require that both circumferential and longitudinal breaks be postulated at Class 1 intermediate break locations 2
chosen on the basis of cumulative usage factor exceeding 0.1.
Justify your positions with, respect to our SRP Section 3.6.2.
110.9 Your response to request 110.3 is not entirely acceptable. Your (3.6.3.2) response indicates that topical report BAW-10132P (March 1977) provides justification for thrust coefficients and break opening times'other than those allowed in SRP Section 3.6.2.
Our review of this report is incomplete. This subject will be considered an open ites pending the results of the topical report review.
110.10 Addit;ional infomation regarding strain rate effects is required (3.6.2) for-Sections 3.6.2.2 and 3.6.3.3:
(3.6.3)
(1) -FSAR Section 3.6.2.2.f states that the minimum specified yield strength'may be increased by up to 20% to~ account for strain rate effects. Our position in SRP Section 3.6.2, Subsection.III.2.a states that the yield' strength may only be increased by 10%. Justify your position in view of your deviation from our position.
(2) FSAR Section 3.6.3.3 does not discuss this subject. Either state that NSSS analyses do not increase the yield strength
- by more than 10*' to account for strain rate effects or
_ provide justification for doing so.
2
~-
9 g
g
+-me v
O
_T
~
110-2 110.11 Identify the computer program to be used for the jet impingement (3.6.3) analysis described in FSAR Section 3.6.3.3 (bottom paragraph of.
FSAR page 3.6-56, Revision 3). Provide ~ verification for, the program in accordance with SRP Section 3.9.1 if it has not already been addressed in FSAR Section 3.9.1.
110.12.
SRP Section 3.6.2, Subsections III.2.b(2) and (3) describe acceptable (3.6.3) methods of analysis for determining the effects of pipe whip.
FSAR Section 3.6.3.3 describes the criteria used in your pipe whip analyses.
(1) -Describe the factor used to account for rebound effects whenever an energy balance analysis is used. Provide justifi-cation for any factor less than 1.2 as required by III.2.b(2) of SRP Section 3.6.2.
(2).. Describe the dynamic load factor used when the restraint is analyzed statically as required by III.2.b(3) of SRP Section 3.6.2.
110.13 Provide the following data for both NSSS and 80P pipe whip
.(3.6.2.2) restraints:
(3.6.3.3)-
(1) The deformation limits for any energy absorbing materials used in pipe whip restraints.
(2) Justification if these limits exceed 50'.' of the ultimate uniform strain.
(3) Force-deformation diagrams for the energy absorbing materials.
Describe how your analyses consider the non-linearity of these diagrams.
(4) Drawings of any pipe whip restraints which utlize energy absorbing' materials.
110.14 FSAR Section 3.9.l.2.1 states that ANSYS was not used for any (3.9.1) nonlinear analysis. FSAR Section 3.9.1.4.1 states that ANSYS is used for elastic-inelastic. analyses.
We require the information requested in our request 110.4 for ANSYS for any elastic-inelastic analyses.
4
n
^
110-3 110.15 Regarding your discussion in FSAR Section 3.9.2.1 on preoperational
-(3.9.2) vibration and themal effects test program for piping:
=(1) Expand the scope of these tests to_ include:all high energy lines and all seismic Category I moderate energy lines.
-(2) Provide a more detailed list of transients that are included in the test program including pump starts and trips, valve closures, etc.
(3) Provide the acceptance criteria against which the measured vibration amplitudes are compared and the bases for these acceptance criteria.
110.16 Provide a list'of mechanical components required for achieving hot (3.9.2.2) standby and cold shutdown of the plants after an SSE for both NSSS and BOP scope. Provide a cross reference to each component's
. seismic qualification summary in FSAR Table 3.9.3-17.
4 110.17 We have reviewed the following FSAR tables for load combinations (3.9.3) and allowable stresses:
A)
BOP Equipment Purchased After July 1,1975 3A.l.48-1 Class 1 Piping and Vessels 3A.1.48-2 Class 1 Valves (both active and inactive) 3A.1.48-3 Class 2, 3 Vessels Class 2, 3 Piping (both active and inactive) 3A.l.48-4 Class 2, 3 pumps 3A.1.48-5 3A.l.a8-6 Class 2, 3 Valves (both active and inactive) 3A.1.48-7' Load Combinations Applied to Tables 3A.l.48-1 to 6 4
3.9-3 Load Combinations Applied to Tables 3.9-6 and 7 3.9-6 Class 1 Component Supports 3.9-7 Class 2, 3 Ccmponent Supports
.B)
NSSS Equipment 3.9-4 Load Combinations and Stress Limits for Class 1 Vessels 3.9-5 Load Combinations and Stress Limits for Class i Piping Additionally, Part I of 3A.l.48 discusses in a general sense the loads considered and the stress limits applied for 50P equipment purchased before July;1, 1975. Part III of 3A.1.48 is referenced by Table 3.9-1 as providing loading combinations and stress limits for NSSS equipment. However, Part III provides no such information.
Also, FSAR Section 3.9.3.4.2 briefly addresses NSSS component supports.
-m-a.
t w
Ms.
+
+
4 P'-'
O
c p
l D
110-4
. Provide the following information so that we may complete our review:
(1) Load combination tables and corresponding allowable stresses for Class 1, 2, and 3 pumps, valves, piping, vessels, and supports in the BCP scope which were purchased before July 1, 1975. These tables should follow the format used in Tables 3A.1.48-1 to 7.
(2) Load combination tables and corresponding allowable stresses for Class 1, 2, and 3 pumps, valves, piping, vessels, and supports in the NSSS scope other than what is addressed in Tables 3.9-4 and 5.
The format used in Tables 3.9-4 and 5 is acceptable.
(3) Modification of Table 3A.1.48-7, Tables 3.9-3 to 5, and any forthcoming Class 1 load combination tables to address the design transients in Table 3.9-2.
FSAR Section 3.9.1.1 has stated that these tranF' ants were considered in the design of all Class 1 items.
110.18 Provide the allowable buckling loads for Class 1 component (3.9.3) supports subjected to faultdd load combinations. Provide justification r-if your criteria exceed the limits of Paragraph F-1370(c) of the ASME Code Section III, Appendix F.
110.19 Your letter of October 6,1977 regarding asymmetric cavity pressuriza-(3.9.3) tion loads endorsed topical report SAI-050-77-PA. This topical (RSP) uses a probability arguement to show that the asymmetric loadings due to a LOCA occurring within the vessel cavity need not be considered in the design of the reactor coolant pressure boundary or its supports.
We have concluded that such probability arguments do.not providee an acceptable basis for long term operation without an assessment of the risk resulting from these postulated transient loads.
You have already committed to combine LOCA + SSE for the design of AS?tE Class 1, 2, and 3 components in FSAR Section 3.9.3.
We will require that, in addition to these' existing commitments, you design the reactor coolant system and its. supports for the following load combinations while limiting the resulting stresses to the faulted allowables as listed in FSAR Section 3.9.3:
y 110-5 Weight + Nomal Operating Loads + SSE + LOCA
- where, LOCA = all effects of an ASME Class 1 pipe rupture including the asynnetric cavity pressurization caused by a break at a reactor vessel, pressurizer, steam generator, or reactor coolant pump
~
nozzle.
It is our position that the peak loads resulting from SSE and LOCA be combined by absolute sum unless acceptable justification is first provided for any alternative method of combination.
110.20 Regarding your response to Request 110.14, we note that footnotes (3.9.3) in Tables 3.9-9 to 15. state that an elastic-plastic analysis was performed to justify exceeding the elastic limit for primary plus secondary stress intensity. Describe this analysis in more detail and specifically address N-417.6(a) (1-3) of the ASME Code Section III (1968).
110.21 We require the following additional infomation regarding your (3.9.3)
Class 1, 2, and 3 system stress summaries:
(1) As required by subsection 3.9.3.1 of Regulatory Guide 1.70, provide a summary of the maximum total stress and defomation values compared to the allowable values for each of the normal, upset, emergency, and faulted conditions for all ASME Class 2 and 3 piping, pumps, valves, supports, and vessels required to achieve. cold shutdown or mitigate the consequences of a postulated pipe break without offsite power.
(2) Provide' the same infomation with the addition of a cumulative usage factor summary for all ASME Class 1 valves and supports.
110.22 Tables 3.9-17, sheet 2/61, Part E, states that an elastic seismic (3.9.3.4)
. analysis was conducted on all components whose stresses due to faulted loads were within 10% of the allowable stress limit.
This implies that some components were not evaluated for SSE loads. Provide and discuss the following:
(1) Provide the faulted load combinations and allowable stresses for which the Hydraulic Shock Suppressors (C-70) were designed.
(2)
Identify and justify any of these components which were not evaluated for SSE. loads.
4 p
- Y
n 110-6 110.23.
We require that all valve operators, and other electrical, mechanical, (3.9.3) pneumatic, or hydraulic appurtenances attached-to active pumps (App.3A) or valves be qualified by test. This requirement was stated by (RSP) letter to you, dated September 24, 1976. Modify FSAR Section 3A.l.48 to describe this test program for NSSS and BOP equipment
-purchased prior to July 1,1975. A program such as described for 809 equipment purchased after July 1, 1975, will be acceptable.
110.24 Regarding your discussion in FSAR Appendix 3A.l.48, we require (3.9.3).
that you modify paragraph (d) at the bottom of FSAR page 3A-72 (App. 3A) to specify IEEE 344-1975.
(RSP)-
110.25 The attached Appendix 110-1 provides guidance for submitting your (3.9.6) initial 20-month inservice testing program of pumps and valves and for requesting relief from the ASME Section XI, IWP, and IWV requirements. Provide assurance that you will submit your initial 20-month program and any relief requests in a timely manner in accordance with this attachment.
110.26 A review of your seismic design adequacy of safety related (3.10) electrical equipment will be perfomed by our Seismic Qualification (3.9.2.2)~
Review Team (SQRT). A site visit at some future date will be necessary to inspect and otheraise evaluate selected equipment after our review of the following requested information. Attached Appendix 110-2 describes SQRT and its procedures. Notice that Section IV.2.A describes information you should submit so that SQRT can perform its review. Attachment 1 thereto provides a standard format for this information. We require this infomation for the equipment described in FSAR Subsections:
3.10.4.1.1 4.16 kV !?etal - Clad Switchgear 3.10.4.1.6 125 Vdc Distribution Centers 3.10.4.1.9 Containment Electric Penetration Assemblies 3.10.4.1.18 tiajor Instrument Package 3.10.4.2 NSSS Equipment (only those required to achieve hot standby and cold shutdown)
.110.27 Regulatory Guide 1.121 describes a method acceptable to us for (App 3A) establishing the limiting safe conditions of tube degradation of
. steam generator tubing, beyond which defective tubes as established by inservice inspection should be removed from service by welding plugs at each end of the tube. Discuss your capability for and intended perfomance in complying with this guide. Justify any alternative criteria you may propose.
110.28 It is cur position that whenever Service Limit B is exceeded, (3*9) areas of structural discontinuity in ASME Class 2 and safety related
. Class 3 piping and thin walled tanks and vessels must be demonstrated
-to retain sufficient dimensional stability at service conditions so as not to impair the component safety function. While inclusien-I
. - ~
p 110-7
~
of. secondary stresses produced by constraint of free end displace-ment is not required to satisfy the stress limits during the emergency or faulted conditions, the reaction loads resulting from the con-straint of free end displacement must be included in the functional capability evaluation.
Demonstrate that areas of structural discontinuity will retain sufficient dimensional' stability to deliver rated flow whenever Service Limits C or 0 are used.
You are also referred to Paragraph NA 2142.2 of the ASME Code that discusses large defomations which are possible in areas Of structural discontinuity stressed to Service Limit C and gross general defor-mations which are possible at Service Limit D.
Although this 4
does not imply that.large deformations will occur in every case where' Service Limit B is exceeded, it is our position that an approach such as the following is to be used:
The analyst should examine areas of structural discontinuity, in the. context of the geometry and stresses in the system in which they exist, to insure that collapse cannot occur at either the equipment nozzles or in _ the piping. Examples of possible collapse modes are situations, such as:
(1) A piping system with a cantilvered length of straight pipe where the fonnation of one hinge would lead to gross plastic defamation, and (2) A piping system with two anchors,'where three points stressed to Service Limits C or 0 could fom hinges and lead to gross l
pastic defomation.
~
If a possible collapse mode is identified, a sufficiently detailed analysis should be perfonned to insure that functional capability is not impaired.
For further explanation of the staff position on Service Limits, operability assurance, and functional capability, see attached Appendix 110-3.
e aW %e -gg,.s p.'
q 110-8 110.29 The steam generator tube wall defects detected during the baseline inspection on Three Mile Island Unit 2 are suspected to have been (5.4.2) ' caused during the fabrication of the tubes, and during installation-of. the tubes in the steam-generators. Describe the precautionary measures considered both during the manufacturing process of the tubes as well as during the fabrication and installation of the steam generators that would prevent the recurrence of such defects in the lildland steam generators.
T M
=g 1
ame e
i
- m. 4 y
- - ~
w m.-
p.-
-y y,
m-m n
g APPENDIX 110-1 v
NRC STAFF CCKiEMTS ON INSERVICE FUMP AND YALVE TE RELIEF REQUESTS The NRC staff', af:er reviewing a nu=cer of pump and valve testing programs, has determined that further ' guidance mign: be hel;ful to illustra e the type and exten: ef inf:rmatien we feel is necessary to ex;ddite die
~
review cf these programs.
We feel that the Licensee can, by incorporating these guidelines into each pr gram submittal, reduce censiderably the staff's review time and time spent by the Licensee in respending to URC staff requests for additienal informatien.
The pump testing program shculd include all safety rela:ed' Class 1, 2 and 2 put;s uhica are installed in water cc: led nuclear p war plants and whien are prov'ided wi:h an emer;~ency power source.
The valve tes:ing program sh uld include all the safe:y related valves in the following systems excluding valves used for opera:ing convenien:e only, such as canual vent, drain, instrument and test valves, and valves used fer maintenance only.
m Pv!R a.
High Pressure Inje::ica System b.
Low Pressure Injection System c.
Accumulator Systems d.
Centainmen: Spray System Primary and Seccndary System Safety and Relief Valves e.
f.
Auxiliary Feedwater Systems
'g.
Reactor Building C oling System h.
Activa Cem;cnents in Service Water and Instr. ment Air Systems which are required to su: pert safety systa= functiers.
i.
Centaincen: Iscla:icn Valves required :c change positien :: isolate containment.
j.
Chemical & Volu:2 Control System k.
Other key c:::enen:s in Auxiliary Systa=3 which are re:uired to direc:iy su scr plant shutd:..n or safe ty system fuh::icn.
- Safety related - ne:esscry to safely shut d:wn the piant and citigate the c:nsequen:cs of an ac:ident.
4 i
s
.'2 1.
. Residual Heat Re:: val Systes m.
Reacter C clan: Systc=
C'n'R a.
High Press:re Core Inje :icn Syst:m b.
L:w Pressure Core Inje::ico System c.
Residual Matt Removal System (Shutd:wn Cooling Systa=)
d.
E=ergency Condenser System (Iscla:f on C ndenter Syste=)
e.
L:w Pressure C:re Spray System f.
Centair.: n: Spriy System g.
Safety, Relief, and Safety /P.elief Valves h.
RCIC (Rea:::r C:re Is:lation Coeling) Syste 1.
Centainment C: cling Systen 4
J.
Centaine:nt isolation velves required to change positi:n to isolate containmen
_k.
Standby liquid : rtr:1 system (3:ren System)
~
1.
Aut::s ic Ce:ressuri:ati:n Sys:c: (any pilot or c:ntr:1 valves, asse:f ata:
hydrauli: Or ;neuma:f:-systems, e :.)
m.
Centr:i R:d Crive Hydrauli: Syste
(" Scram" functibn) 6:cer key :::::nenis in Auxiliary Systems which are requi, red t: dire::1;.
n.
~
su:: rt plar.: snutd:,.r. Or safo:y :ys:em fun :i n.
o.
React:r C: lant System Instervice Pu and Valve Testir.; Pree am I.
Inf:rma'.ica re uf red for ::RC Staff Review cf the Pump and Vaive Testing Pr: gram A.
Three sets cf P&ID's, which include eil cf the systems listed above, with the c de class and system becndaries :learly marked.
The drawin;s shculd inclu: all of the ::=;:nen:s ;resan at :ne time of submi :al and a le:end of ::4e FLIC sy :cis.
B.
Identificatica cf :ne applica:ie AS"E C :e Edi:icn and Adder.:;
C.
The period for which the progra is applicable.
D.
Identify the c =ponent c:de class.
1 P
1 1
W
... - - - ~ _
E.
F:r Pump testing:
Identify 1.
Each pump rac. aired t: be tested (nano and number) 2.
The test para aters t: be cassursd 3.
The test frecuency F.
For valve testing:
Identify 1.
Each valve in ASME Secti:n XI Cate; rie: A & 5 that will be exercised every three : nths during ner=al plant operatien (indicate whe:ner partial Or full streke exercise, and for p:wer cpera:ad valves list the 11 iting valua for str:ketim2.)
2.
Each valve in ASME See:f en XI Categ:ry A that will be leak,
tested during refueling Outages (Indicate the leak tes:
precedure y:u intend :: use) 3.
Each valve in ASME Section X* Cates:rias C, O and E that will be tested, the type Of tes and the test frequency.
For check vaivas, iden:ify th:se that will be exer:ised every 3 m:nths and th:se that will cnly be exercised during cold shutde,.n Or refuelin; cuteges.
II. Additi nal Inf:reati:n That 'Will Be Hel;ful in Speeding U; the Review Pr: cess A.'
Include the valve 1::ati:n :: Ordinates or ::her a:;repria:e locatica infer:.,a:ica whi:h will ex;edite cur locating th:
valves on the P& ids.
B.
Provide' P&ID drawings that are large and clear en: ugh to be read easily.
C.
Identify valves that are pr:vided with an intericek :: other c::::nents and a brief d:scri;;ien of that functi:n.
Relief Reevests fr:: Se:tien XI Re uiecments The larges; area of cor.cern fcr the ? RO s:sff, in the review of an inservica valve and pum; testin; ;r: gram, is,in evaluating tha basis for justifying r lici fre: Secti:n XI Require:en:s.
IthasbaenOurex;irie$ce e
.e ai v
y
4 that many requests f:r relief, submitted in these pr: grams, de net pr: vide adequate descriptive and detailed technical informati:n. This e'xpiteit informatien is necessary :: provide reascnable assurance that the burden imp: sed en the licensee in c::;1ying with the code requirteents is n:t justified by the increased level of safety ct:ained.
~
Relief reques:s which are submitted with a justificatien such as "I= practical", "Inae:essible", or any other categ:rical basis, will requirt additional infon=atien, as illustrated in the enciesed examples, := alle, cur staff t: make an evaluatten of that relief request. The intentien of this guidants is to illustrate the centen: and extent of informatitn rs;ut.ed by the NRC staff, in the reques: for relief, :: take a preper evaluatien dad adequately d:Oument the basis fer tha relief in cur safety evaluatiCn rep rt. The N?O staff fasis that by receiving this inf:rtati:n in tne pr:gra: submittal, subsequen: requests f:r additi:nal inf:rmati n and delays in c pleting our review can be considara:ly reduced er eliminated.
s I.
Inf:rration Recu_i_ rad for NRC review of Relief Rect:sts A.
Identify cc:::nent f:r unica relief is reques:ed:
1.
Name and nur.ter as given in FSAR 2.
Functica 3.
ASME Secti:n III Code Class a.
For valve testing, aise s;e:ify the ASMI Se::icn XI valve categ:ry as defined in I'4V-2000 G,
5:ecifically identify the ASME C:de rt;ui,rement that has been determined to be impractical for each c: ;cnent.
C.
Previde inf:rmati n :: su:;ert the determination that the reeuircren; in (3) is i=:racti:al; i.e., state and ex; lain th.e basis'for recuesting relief.
D.
- ipecify :ne inservice :ss:ing :nat will :e ;erferred in lieu of the ASME C:de See:i:n XI requirer.sn:s.
E.
Provide tne schedule fcr it;lementatica of the precedure(:)
s in (0).
w
g e
m.
II. Examples to 111ustrat: Several 70ss'ible Areas 'Where Ralief May Ee Granted and the Extent and Centen: cf Informati:n Necessary to Make An Evaluatica A.
Accessi$ fifty: The regulation spa:f fic:11y grants relief frca the cede re;uirement because of insufficient ac:ess pro-visicas. H: wever, a detailed discussi:n of actual physical arrangement of the c: ;cnent in questien to illustrate the insufficiency of space for c:nducting the required test is necessary.
Discuss in detaff the physical arrangement of the c::penent
_in questien to demonstrate that t.P.ere is not sufficient space
- t: perform the c:de required irservice testing.
What alternative surveillance means which will pr vide an' acceptable level of safety have you :ensidered and why are these. eans not feasible?
B.
Envirencental Conditions (e.g., High radiatien 1svel, High
~-
tempercture, High humidity, etc.)
Although it is prudent to maintain oc:u;atien radiation ex;;sure fer'inspec:icn pers:nnel as icw as practicable, the request for relief frem the c:de requirteents cannet be granted sclely en the basis of high radiatien levels alene. A balanced f,udgment
~
between the hard: hips and :::;ensating increase in the level of safety should be carefully established.
If the health any safety of the public di:tates the necessity of inservice testing,siternative reans or even decentaminatica of tne plant if necessary shculd be pr:vided er deveic;ad.
Pre' vide additicnal ini:rmatien regarding the radiation levels at the required tes: Iccatica.
'4 hat alternative testing te:nni;ues-which will pr: vide an ac:eptable level of assurance of the integrity of the c:::enent in questier. have you ::nsidered and why are these techniques determined :: be i ;ractical?
.w.
4
-~
~ ~ '
- ~
r,
~_
6-a C.
Instruman:ation is net originally pr:vided Provide in~0nr.sti:n to justify that c::pliance with 'the c:de requiremen:: would resul in undue burden or hardships withcu:
a c:r?ensating inc. ease in the level of plant safety. ' Wha t alternative testing mathods which will pr: vide an acce;;asle level of safety have ycu censidered and why are these metheds determined to be impractical?
D.
Valve Cycling Ouring Plan: Opera:icn Ceuld Put the Plan: in an Unsafe Conditicn The licensee sh uld explain in detail why exar:ising de::s -
during plant 0;eration could je ;ardi:e the piant safety.
E.
Valve Testing at C:1d Shutican er Refueling Intervals :n lieu of the 3 M n h Ecquired Interval The licenses shcuic ex: lain in de: ail why each valve cannc: be exercised during n:rmal cperation. Also, f r the valves where a refueling interval is indica:e4 explain in detail why each
' ~
valve cannet be exercised during cold shutd:wn intervals.
- III. A :s::ance Criteria for Relief R?:ue_s:
The Liter.see must s6c:essfully dem:nstrate that:
1.
Complicnce wi:n the cede requirements w:uld result in hardships cr unusuki difficulties without a cc :ensating increase in the level cf safety and cone: ;iiance will provide an ac:e;;abic level of cuality and safety, or 2.
Fr: posed ai ernatives to the c:de requiremen:s or ;0rti ns th re:f will ;r: vide an ac:eptacle level cf quali:y and safety.
Standard Forma:_
A standard f:rmat, for the valve ;Jrtien of the ;; : and valve testing program and relief rc: ests, is included as an atta n=ent :: this Guidar.:e.
s The NEC staff believes :na; :nis standard forma: will redu 2 the time s;ep; by both the s;aff in our review and by the licensee in their preparation
e,
m
_ '7 of' the pump and valve testing progran and submittals. The standard format includes exam:las of relief requests whien are intended t: illustrate the cppli:ation Of the standard f:rmat and are not necessarily a specific plan relief request.
O F
' O e
O e
6 M
AM, ~A me e'J..=.W*--
s
~.
TO APPENDIX 110-1 STANDARD FCR."AT VALVE INS RVI C......s' P:Cr"w"I ("v':I"~' ~'"
- . i n i tii 6
4 0
0 L*
O O
S e
6
e 1___.__,_ y U
n' 9
3 REl4 ARKS a8
- l' h
,l,)
(Not to be used for relief basis) i
'D O
[
U E
T.
0 l
E 0;
i e
a a',
i la valve C
S a,
'l
'L valve U
E s
E T.
1:uinber 3 T TT U T m-3 jj ;y,yI y',E Cater:ory as o
l
)
710 3
0-14 X
4 GA H
Lo tr i
I 700 3
0-15 X
11T 717 3
C-IS X
16 CK SA CV X
CS 702C 3
C-15 X
16 CK SA CV 707 3
E-14 X
3 ftEL SA CV i
034 3
0-11 X
X 4
GL C
Q X
!!I 60 sec.
I 722B 3
0-11 X
B-ll X
A-10 X
3 REL SA SitV 729 2
B-10 X
'e 744B 2
D-14 X
10 GA i40 C
Q 3
. LT '
X
~
ilf 30 sec.
i 1
[
4 0
1 d
4 t
's
4hu.'e eN"-'4+
'a7 m
~_
Lerend for Valve Testine Exa :cle For.at Q
Exercise valve (full strche) for operability every (3) 5.cn:hs
~
LT - Yalves are leak tested per Secticn XI Article IW-3420 MT
-Stroke tica measurements are taken and concared to the strcke tire limiting value per Section XI Article IW 3410 CV - Exercise chec': valves to the pcsition recuired to fulfill their functica overy (3) =cnths SRV-Screty and relief valves are tested per Sactica XI Article IW-3510 DT - Test category 0 valves par Section XI Article IW-3c00 ET - Ycrify and record valve positten before c;erations are performed and after c;erations cre cc=pleted, and verify that valve is lecked or sealed.
CS - Exercise valve for cperability every cold s5utdown RR - Exercise valve fcr coerabili:y every reactor refueling
. %e=.
D a
I 1
..-. ~..
,=...
-l r.
Relief Request Basis Sys tem: Auxiliary Coolant System, Component Cooling 1.
Valve:
717 Category:
C
' Class:
3 Function:
Prevent backflow fran the reactor ccolant pump cooling coils Impractical test requirement: Exercise valve for operability every three months Basis for relief: To test this valve would require inter uption of cooling water to the reactor ccolant pumps motor cooling coils. This action could result in damage to the reactor ccolant pumps and thus place' the plant in an unsafe ode of operation.
Alternative This valve will be exercised for operability Testing:
during cold shutdowns.
2.
Valve:
834 s
Category:
B-E Class:
3 Functicn:
Isolate the primary water frem the ccmcenent cooling surge tank during plant operation.
It is normally in the closed pcsitien, but routine operation of this valve will cccur during refueling and cold shutdowns.
Impractical Test Exercise valve (full stroke) for operability
-I Requirement:
every th ree (3) mon ths.
Basis for Relief: This valve is not required to change position during plant operation to acccmplish its safety function. Exercising this valve will increase the possibility of surge tank linii contaminaticn.
?
l Alternate y Verify and record valve positicn before and
~
Testing:
and af ter each valve c;eration, j
I
, see w e.
m--m,.-...
,u-,.
~
~ /s/ ~.
.~
-2 3.
Yalve:
7443 Categcry:
A Class:
2 Functicn:
Isciate the nsideal heat exchangces frca the celd leg R.C.S. backficw and c::umulatier backfiew.
~
Test Requirements: Seat leakage test 3 asis fcr This valve is located in a high radiation field Relief:
(2000 mr/hd whien would make the required seat leakage test ha:ardcus to test persennel. We intend to seat leak te: two cther valves (3753 and 3753) wnich are in series with this Yalva and v;ill alsc preven: b e.ckfi cw. We feel' that by cc=plyin; with the sea leakage neutre ents we will not achieve a c:. ;ensatcry incnase in the level ef safety.
Altemative No alternative sea leak testing is proposed..
Tes ting:
8 9
J 0
t m
T
^
WORKING PAPER 2/17/73 Aopendix 110-2 Reg.4 SEISMIC QUALIFICATION RE7IT4 TEAM (SQRT)
I.'
SCOPE SQRT tasks include both generic and si:e specific reviews. Generic reviews cover equipnen: supplied by NSSS and A/E com:non to more than one plant. Specific plant reviews as delineated in the Standard Reviev Plans, Section 3.10 vill be supple =ented by.SQRT site visits and evaluation.
II.
CBJECTIVES SQRT is a g cup of NRC staff ne=bers established to condue: reviews of the seis=ic desip adequacy of saf e:7 related sechanical components, electrical ins::u:nenta: ion and their supper:ing structures :o acco=plish the following:
1.
Changes in seismic qualification cri:eria, such as the revision of IEEE-344 Standard and the issuance of Regula:ory Guide 1.100, require that the staff verif :
7 (a) For older plan:s having cc ponents qualified unde: previous criteria; that co=penents have adequate =argin to perfor=
their intended design func:icas during a seis=ic event.
(') For new plant applicaticss; that there has been unifor=ity and consistency in i=ple=enting the new criteria.
2.
Deter =ine the desip adequacy of selected cc=penents for seis=ic loading condi: ions.
1 i
l o
e' "ew
-H=-
O
-)
In the case of plants which have design basis *eiscie ground 3.
motion levels increased, reviev to assure adequate design margin exists a: the revised levels.
3 III. CENERAI, PROCEDURES SQRT will conduce generic and plant specific revievs:
Generic reviews vill be conducted of all NSSS vendors and wat 1.
architect engineers (=ajor equipment venders and tes:ing laboratories may be included if ne: essary) to assu e proper interpretation and i=plementation of current sais=ic qualification criteria applied to plants applying for construction per=its and opera:ing licenses.
A plant specific saissic review will be conducted of each plant cov 2.
undergoing licensing review having cc:ponents qualified to the IEEE-344, 1971 criteria.
A.
For ce=pesents havi g =ulti-plant application, (such as chose within the scope of an NSSS vender) seismic qualification review a
specific sites vill provide ge:eric i= formation.
at B.
For ce=ponents which have only specific plant application (mostly those within the scope of AE supply) seismic qualification review at specific sites vill provide information for the si:e.
1 3.
Seis=ic qualification reviets for plants with revised increased design basis seis=ic ground =otion levels vill be conducted on a plan: by plan basis.
I I
l
.a.,
"'g
.m
=
3-IV.
S?ECIFIC PROCbCRIS SQRT procedures provide for both generic discussien neecings and plant site visits.
1.
Generic Discussica Meeting:
To implement the generic review specified S III.1 and III.2.A, a generic discussion meeting will be held to discuss the following:
A.
Meeting Agenda Meeting Objectives by SQRI 3.
NSSS or A/E personnel should be prepared to present the following infor=ation:
(1) A detailed dascription of current practice followed in seismic qualification, including acceptance criteria, nethods and-,,
procedures used in conducting testing and analysis.
Present and discuss the seismic qualificatica progran on f
certain specified items (i.e. pumps, valves, diesel l
1 generators, notors, bistable units, relays, etc.)
l (2) Information regarding administrative control of component seismic qualification, especially the handling of interface problems, documentation and internal review procedures.
(3) Identifying the scope of their suppliers. A list of equip-
.l nent should be made available if possible prior to the nesting.
e
--r w
.m 4
,,..-y-
~
.:=
' N
/
,4 4-a 9
For the cases specified in III.2. A., =ethods and procedures for C.
conducting seis=ic qualification review are discussed, including selection of plants f or site visit and setting up a tentative schedule for such visits.
Discuss necessary docunentation.
D.
Testing capability, for=at Inspect testing facilities, if any.
E.
of testing reports, vave for=s of shake table =otions, =enitoring and.;cntrol devices are the sajor ite=s for inspection.
SQRT concludes the =eeting and specifies the follow-up ice =s.
F.
2.
Plant 5:.te Revievs:
To i=plement plant specific seismic qualification reviews specified in III.2 above, on-site inspection of equipment and supporting structures in question is required. Site visits generally follov
- the following procedures:
Pre-visit infor=ation submission:
A.
(plant owner) receives initial information The applicant concerning the intended visit, and should subsequently submit d
two sum ary equipment lists (one for NSSS supplied equipment an one for A/E supplied equipment). In the lists, the following infor=ation should be specified for each iten of equip =ent:
O e
"*-m.mau..a #,
.r, gr rw
=P
e (1). Method of qualification used:
(a) Analysis or test (b) If by test, describe whether it was a single or multi-frequency cust and whether input was single or bi-axial (c) If by analysis, describe whether statie or dynamic, single or sultiple-axis analysis was used. Present natural frequency of equipment.
(2) Indicate whether the equiccent is required for:
(A) hot stand-by and cold shutdewn (3) hot stand-by (C) cold shutdewn (3) Availability for inspection (Is the equipment already installed at the plant site?)
SORT screens the above information and decides which items will be evaluated during our' f orthcoming site visit. Ihe applicant will be infor=ed of these ite=s and will be expected to submit two weeks prior to the visit a seismic qualification su= mary as shown in the Attachment i for each of the selected ite=s.
3.
A brief =eeting is held at the beginning of a site visit with the following agenda:
~
~
(1) SQRT explains the objectives of the site visit and procedures
~
to conduct equipment inspection.
(2) Utility personnel or their designees present an over view of the seis=ic qualification program conducted.
(3) The seismic qualification of certain specified items may be discussed as necessary.
(4) SQRT specifies items that need to be inspected.
C.
SQRT conduct inspection on specified ite=s.
D.
SQRT describes findings of the inspection.
E.
General discussion.
7.-
SQRT concludes the visit and specifies needed information and the follow-up actions.
3.
Plant site revdevs fer cases involving increased design basis seismic g:cund motion.
(under development]
In general utility vill provide data on systems and components used to bring the plant to shutdown and saintain it in a cold shutdown i
condition, Safety cargin for seismic qualification of equipnent should be assessed.
4.
Af ter each visit SQRT will issue' a trip report, which identifies findings, conclusions and follev-up itens. Status reports =ay be issued'as =ecessary. The site review vill include the issuance of an Ivaluation Report for the specific plant. Generic evaluations vill be referenced to the NSSS vender or A/E.
%e' 6,
I O
1
+
e--n
---s----
-3 V.
RE5FONSI3ILITIES OF NRC ?ARTICIPANTS:
The Seismic Qualifica: ion Review Tea = consists of nesbers of A.
the Mechanical Engineering 3 ranch (MI3), :he Ins::usentation and Centrol Sys:e=s 3 ranch (ICS3), and the Plant Systa=s 3 ranch (PS3). One additional member f rom MI3 vill join the :ea=
when a review of a specific plan is going to be conducted. Ihis se=ber vill be the reviewer of the plan:.
The Tea = Leader is responsible for scheduling actions, coordi::ating staff posi: ions and contac:ing vi h appropriate authorities for vc k 4
assign =ents to each member. He reports :o the MI3 Chief regarding the progress of SQRT perf or=ance. He vill set up necessary contacts for generic reviews and vill contact project =anage=ent for specific plan site visits. He vill specify the see:ing ebjectives and concludes :eetings.
n e MI3 = embers and Tea: Leader are responsible for reviewing assigned seismic qualifications in :he area of responsibili:y of Mechanical Engineering 3 ranch, including :he =e:heds and procedures used in test and analysis.
Me=bers representing the Fever Syste=s 3 ranch (PSB) and :he Ins:rumen:ation & Con ol Syste~.s 3 ranch (ICS3) are respensible for reviewing assigned seismic qualification in the area of their branch, including ecuipment signal interpreta: ions fec functional terification 3ey serve as a liaison between SQRT and ICS3 and PS3.
J l
~..- -
n
- p All me=bers shall present : heir opinion and professional judgement and uniform to the Team Leader in order to arrire at consistent SQRT pesi: ions.
The ME3, PS3, and ICS3 project reviewers will be advised of -SQRT 3.
ac:ivities which relate :o specific plants. The Franch preject reviewer is responsible for evalua:ing the impact of SQRT. activity on the specific plant review and for taking appropriate action to The include pertinent information in the plant saf ety evaluation.
branch projec: reviewer is expec:ed to participa:e in the site visi:
The ME3 reviaver and attend pertinent generic neetings as necessary.
will have further responsibilities in those cases where revised seismic loads have been es:ablished.
-s The DFM project manager, after being infor=ed of the intended plant visit, is expected to contac: :he applicant and arrange for the visit. The project manager serves as a liaison between the SQRT and the applicant.
o C.
Generic =eetings will be arranged by the SQRT or via che DPM generic project manager if one is assigned.
Reprasentatives from I & E Regional Offices and other interested-D.
organi:ational groups within NRC are vele:ce to attend either The SQRT should l
generic nee:ings or plant site visits as observers.
be infor=ed of expec:ed at:endance at such reetings or site visits.
5
-g-
--g--
-y 4
,+m4 y
-p w
n-u-y,
. - -. ~ -
,4
]-
=~
ATTACHMENT 1 to APPENDIX 110-2 Seismic Oualification Smry of Equipment I.
Plant Nane:
Ty;e:
P'a1 1.
Utility:
2.
NSSS:
B'al 3.
A-E:
II.
Component Name
..od el v
Cuantit71 1.
Nu=ber 2.
Vendor 3.
Physical Description 4.
Location: Building:
(In Plant) 31,y,ggag 5.
Natural 7:equencies in Each Direction:
)
6.
Functional
Description:
7.
Pe:Cinent Reference Design Specifications:
III.
Is Ecuipment Available for Inscection in the Plant:( } Tes [ } No Co==ents:
l l
A m
. - e IV.
Seismic Oualification Method: Test:
Analysis:
Combination of Test and Analysis:
V.
Seism!.c Input:
1.
Required Response Spectra (attach the graphs):
Required 2.
Acceleration in each Direction:
VI.
If Cualification by Test, thenConclete:
1.( ] Single Frequency
(] Multi-Frequency 2.[ ] Single Axis
[] Multi-Axis 3.
Frequency Range:
4 TRS enveloping RRS using Multi-Frequency Test [] Yes(attachTRSgraphs)j
(] No 7=
5.
g-level Test at g=
h
=
3 h
V" 6.
g-level Required g=
2" 7.
Mounting:
1.
Seismic Report:
2.
Field Check:
S.
Functional Verification Perforned( ] Yes [ ] No ( ] Not Applicable VII. If Qualification by Analysis or by the Combination of Test and Analysis then, Complete 1.
Description of Test including Results:
y w
ya g
--,-m--
y ap,-
9 n-
n-m.
-e.
.s
.a.
2.
Method of Analysis:
[ ] Static Analysis ( } Equivalent Static Analysis [ } Dyr.2mic Analysis
[ ] Res'ponse Spectrum
( l Time-History 3.
Model Type (each direction):
4.
Computer Codes:
5.
Damping:
6.
Support Considerations:
7.
Critical Structural Elements:
Governing Seismic Total Stress A.
Identification Location Response Combinatien Stress Stress Allevable
~
Effect UPon Functional 3.
Max. Deflectics Location Ocerabilitv I
l k
l
/
APPENDIX 110-3~
2 Staff Guidance for Essential PipingsPumps and valves I.
In applying the cri:eria-stated belov the following prerequisites are to be applied:
(1) All Code require =ents for ASME Code Class components must 1Ht satisfied. In particular the Code pressure limits specified in conjunction with St evice Stress Limi:s cannot be exceeded (Refer 4
to NC-3000). Code design limits sust be satisfied for those leads specified as design.
(2) 'Jhcn a Service Stress Li=it above 3 is specified both operability and functional capability =ust be de=onstrated for that limit.
In addition active pumps and valve operability sust be de=enstrated
. irrespective of stress limit.
II.
Definitions:
Active Cemoonen: - A pump or valve relied upon to shut down the plant or to prevent or si:igate the consequences of an accident.
Component and Supeor: Functienal Cacability - Ability of a component, including its supports, in a Safety Related System to deliver rated flow and retain dimensional stability when the design and service leads, and their resulting stresses and strains, are at prescribed levels.
Component and Support Operabiliev_ - Ability of an ac:ive component, including its supports, in a Safety Related Systes to - perform the mechanical notion required to fulfill its designated saf ety func: ion 6*hm=s w
r-
- < w e y.
s
. a
.when the design and service loads, and their resul:ing stresses and strains, are at prescribed levels.
Essential Systems - Any of the following:
(1) Any AEMI Code Class 2 systes, or (2) Those ASME Code Class 3 systems which perform a safety related function.
These are :he functional syste=s necessary to assure:
(1) The capability to shut down the reactor and =aintain 1: in a saf e shutdown condition, or (2) The capability to prevent or sitigate the consequences of acciden:s which could result in potential offsite exposures comparable to the guideline exposures of 10 CFR 100.
III. Service iisits above "3" for Essential Systens Service Limit C The use of service limit C for components and component supports of essential systens in lieu of Level 3, nay be persi::ed for designated
]
components and component supports provided that (1) the operability of active pumps and valves is demonstrated at service limi C; (2) the functional capability require =ents at service limit C are satisfied; and (3) constraint of free end displacemen: has been considered for the specified loadings.
4
c.
,e Sersice L1=it D The use of service lini: D for components and component supports of essential systems in lieu of seriice lisi: 3, say be permitted for designated components and component supports provided that (1) an analysis comparable to that described in N3-3200 of the Code is conducted which will desonstrate that the designated service loading combination will not produce stresses greater than Level D, and supplemented by additional analyses that show that resulting strains and deflections vill not impair the ability of the system to operate normally for extended periods of :i=e; (2) che operability of active pumps and valves is demonstra:ed at service lisi D; (3) the functional capability requirements at sertice 11=1 D are satiefied; and (4) cons:raint of free end displace =ent has been considered for the specified loadings, i
IV.
Operabilit_v and Tunctional Cacabiliev Active Puses and Valves l
The design of active pu=ps and valves including their supports, which must perfors a mechanical =ction to fulfill its safe:7 related function may utili:e any of the four service limits including service l
limits C and D, provided an operability assurance program (meeting i
SRP 3.9.3.Section II.2) demons: rates tha: :he pu=p or valve as supported O
can adequately sustain the designated combined service loadings at a s:ress level at least equal :o the specified service limit, and can perfors its safety function without i= pair =ent.
1P
"*-****2"*
N-
..e....
.m..
r.
s 4-e 3
The operability require =ents specified for =echanical and hydraulic snubbers installed on safety related systems is subject to review by the staff. klen snubbers are used, their need shall be clearly established and their design criteria presented.
Functional' Capability Areas of s:ructural discontinuity in piping, tank.s and vessels (such as piping elbows, teen connections, tank and vessel nozzle areas, and thin-valled tanks) shall meet service 11=1: 3, but alternatively, =ay be permit:ed to designate service limits C or D, provided it is deson-strated that the disecutinuity areas retain sufficient d1=ensional stability at service conditions so as nec to impair the component functional capability. Justification shall be provided by tests, or analysis or combina: ions thereof.
Constraint of Free 2nd Displacement Component and supper: loads produced by the constraint of free end displacement resulting frc= thernal or other movement (such as anchor pois: movement) shall be evaluated for compliance w1:h Code specified stress li=its and shall also be included in the operability assurance program for active pumps and valves, and in the fune:1onal capability evaluation for areas of structural discontinuity. % tile inclusion of these loads to satisfy service stress limits when level C or level D stress li=its are designa:ed is not required, the reaction loads resulting from :he constraint of free and displacement shall be included in the operability assurance progras and the functional
+w i.
)
-S-capability evaluation for any level of stress limit which =ay be designated, including levels C and D.
In each case where the categorization of stresses produced by the constraint of free end displacement is nade, the consideration of these stresses and the service limits as secondary rather than pri=ary shall be justified.
l
.m i
4 amns G
E i
121-1
-.~.
121.0 MATERIALS ENGINEERING 3 RANCH - MATERIALS INTEGRITY SECTION 12i.;0 In reference to request 121.5, we require the following (5.3.1) additional information on Midland Plant, Uni Nos. I and 2 reactor vessels:
(1) A schematic sketch of each reactor vessel showing all welds (longitudinal and circumferential), plates and/or forgings in the beltline region. Welds should be identified by a shop control number (such as a procedure qualification number), the heat of filler metal, type and batch of flux, etc. Each plate and forging should be identified by a heat number and material type.
(2) For each of the above welds, and for weids in the vessel material surveillance program, an identification of the welding process should be provided.
(3) A listing of the folicwing infernation en all beltline materials (weld, plate and/or forging): chemical composition (particularly Cu, P, and S content), drop weight T USE and tensile properties.
(If any of these'yh7, RT' T,oughness requirements have not been ractuk t determined, use Branch Technical Position - MTE3 5-2 to estimate their value.)
~!
(4) The maximum end of life fluence at the vessel I.D. and 1/4t locations for each weld in the beltline region.
121.11 Clarify the discrepancies between Table 5.2-3 and Table 5.3-2, (5.2.3.1) with specific reference to materials of construction. Also (5.3.1) clarify the discrepancies between Table 5.3-1 and Table 5.3-2, with specific reference to USE and RT of weld deposits.
NOT 121.12 References to Topical Reports BAW-iC046P and BAW-lC046 are not (5.2.3.3) appropriate. BAW-10046A, for calculation of pre-and post-(5.3.1) irradiation properties of reactor vessel materials (using the (3A l.99)
NRC staff recomendations attached therein), is an acceotable (5.3.2) reference. Amend Section 3A.l.99 to reflect the staff recommendations.
121.13 Clarify the discrepancies that exist betseen Sections 3A1.14 and (3A.1.la) 5.4.1.7 with reference to pump flywheel integrity.
(5.4.1.7)
' ~#j 121 2 121.14-Response to request 121.1 is not adequate. Confirm that 100%
(5.3.5) of all Class 1 components will be given a preservice inspection (5.4.1) as defined in article IWB-21CO of ASME Code Section XI, and Class (6.6) 2 and 3 components will be examined to the extent practical. Parts (16.0)
.1 and 2 of request-121.1 must be fully answered prior to issuance of the OL. Confirm that components subject to inservice inspection, under the applicable rules of the ASME Code that conform to the requirements of 10 CFR Part 50, paragraph 50.55a(g),
shall be fully accessible and inspectable. Any exceptions or deviations should be identified along with complete technical justifications for such exceptions and any alternate provisions that may be proposed in lieu of the applicable requi rements.
121.15 Provide information on the fracture toughness characteristics (3.9.3.4) of primary components supports structures and the minimum operating temperature of these supports.
4 121.16 Response to request 121.2 is not adequate.
It is our (6.6) position that augmented inservice inspection is required.
(3.6.2.1)
Refer to request 110.7.
(RSP) b w
=
't i
122 -
122.0 MATERIALS ENGINEERING BRANCH - METALLURGY SFCTION i-122.1 Clarify your discussion of compliance with Paragraph C4 of (App. 3A,.
of Low-Alloy Steel." Paragraph C.4 states that unless all of the Regulatory Guide 1.50, "Centrol of Preheat Temperature for Welding regulatory positions as stated in Paragraphs C.1, C.2, and C.3 of the guide are met, the weld is sutQect to rejection unless the soundness is' verified by an acceptabie examination procedure.
'Is it your intent to comply with paragraph C.4 of RG 1.50 when the Regulatory Positions Cl, C2, and C3 of RG 1.50 are not met?
122.2 Information on the reactor internals has been supplied for only the reactor core support assembly components. Section 4.5.2 of the Standard Review Plan states that the internals for a pressurized water reactor typically consist of the following structures and components: (1) the lower core support structures. including the core barrel, neutron shield pad assembly, core baffle, lower core plate, and core supports; (2) the upper core support structures, including the top support plate, beam sections, upper core plate,'
support columns, and guide tube assen611es; and (3) the in-core instrumentation support structure. For Sections 4.5.2.1 to 4.5.2.5 of the '
FSAR, provide the infomation requested in Revision 2 of Regulatory Guide 1.70.
~.-
122.3
- Provide the weld metal material specifications for the reactor
-(4.5.2.1) core support assembly components.
122.4 For the reactor core' support assembly components, discuss the (4.5.2.2) degree of compliance with the acceptance criteria of Article NG-5000, " Examination," of the ASME Code.Section III.
122.5 Provide a summary of the acceptance criteria and the fracture (5.2.3.3.1) toughness data for the Class 1 ferritic materials of the steam generators, pressurizer, piping, pumps, and valves of the reactor coolant pressure boundary.
122.6 Recent operating experience with some Babcock and Wilcox once-(5.4.2.1) through steam generators has revealed areas of steam generator tube degradation in the fom of circumferential cracks. Expand your discussion in Section 5.4.2.1 of the FSAR to address the actions taken by the Midland Plant, Unit Nos.1 and 2, to preclude degradation in the steam generators. Discuss the improvements made to prevent inleakage'to steam generators from sources such as-the condensers.
Discuss provisions for access openings to inspect for tube degradation and discuss'other steps taken to facilitate steam generator inspection.
E l c
w.
m:-
122-2
' ~#
122.7 Indicate the degree of conformance with Branch Technical (5.4.2.1.3)
Position MTEB 5-3, " Monitoring of Secondary Side Wawer Chemistry (10.3.5) in PWR Steam Generators," which is appended to Section 5.4.2.1 of the Standard Review Plan.
122.8 Provide the materials specifications for the principal materials (6.3.2.4) of construction for ECCS components within B&W's scope of supply.
122.9 Recent operating experience has indicated deficiencies in the (5.4.2.2-techniques and procedures for plugging steam generator tubes.
In 16,3A) some cases, plugs which were to have been installed were apparently omitted. Poorly installed, leaky plugs have also been experienced at some plants.
In one instance, a steam generator tube plug was found in the reactor vessel.
~
Describe in detail the plugging technique which would be used for the Midland Plant, Units 1 and 2 should this be necessary because of leaking tubes during plant life.
Include a description of any tests of the tube after plugging and your procedure to assure accountability of plugs and plugged tubes.
.. ~ - -
.m.
b 4
6 6
+****~h Ph e
eu,
,pa@
PN w<
.g g*
=
A 130-1,2.-
+
130.0 STRUCTURAL ENGINEERING BRANCH 130.6 Provide a table summarizing the wall and roof thicknesses and the (3.5.3) strengths, including the age specified for each tornado missile barrier.
130.7 Demonstrate with an example that the use of the square-root-sum (3.7.2.7) or-the squares method for closely spaced modes as opposed to the use of the procedures described in Regulatory Guide 1.92 does not have a significant impact on the Midland piping design.
130.8 Section 3.7.2.8 indicates that non-seismic Category I structures (3.7.2.8) are analyzed and designed to prevent failure under SSE conditions.
Describe the method of analysis, the load combinations, and the allowable stresses considered in designing these non-seismic Category I structures.
130.9 Section 3.7.3.2.1 states that the design of structures and the majority-(3.7.3.2) of the equipment is not fatigue controlled. Justify this statement in light of our SRP 3.7.3 position of postulating one SSE and five OBEs during the plant life. The number of cycles per earthquake should be obtained from the synthetic time history (with a minimum duration of 10 seconds) used for the system analysis, or a minimum of 10 maximum stress cycles per earthquake may be assumed.
130.10 As a result of our Regulatcry Guide Review Evaluation for Midland (3.7.4.1) in 1976, it was our understanding that the only exception made to Regulatory Guide 1.12 was the use of response spectrum analyzers in lieu of discrete response spectrum recorders. In the FSAR, many exceptions to Regulatory Guide 1.12 are indicated. Some of these exceptions include using peak strain gauges in lieu of peak recording accelerographs as well as many exceptions to ANSI N 18.5. We require justification and clarification of thase conflicting statements.
130.11-Describe the specific methods vsaJ in the design to account for the (3.8.1) radial tensile forces in the dcre and walls of tha containment.
130.12 Indicate if the value of fc stated in Table 3.5-4 is used in all (3.8.1.6) analysis required by 3.8.1 and if so, indicate the age at which the compressive strength of concrete is specified.
130.13 Your load combination equations and method of analysis in your FSAR (3.8.1)
Section 3.8.l' deviate from our position in ACI 359 and Section 3.8.1 of the SRP. Demonstrate that the degree of conservatism used in the Midland design is equivalent to that which would have resulted if ACI 359 and the SRP had been used.
If your approach is less conservative, provide your basis for concluding that adequate margins of safety exist.
e
~- -
w. ;&.. -
130-2 130.14 You reference the ACI-318-71 Code and ACI 318-63 specifications
~~
(3.8.1) as the major specifications for concrete work. Specifications for the containment concrete and other materials of construction currently acceptable to the staff are listed in the ACI-359 Code with the exceptions specified in Standard Review Plan Section 3.8.1.II.6(a).
Comparisons of your referenced documents reveal several deviations from the staff position. To enable us to evaluate compatibility between the two sets, provide a list of the specifications used for Midland parallel to those listed in the ACI-359 Code. This information should provide for an evaluation and justification of these deviations from the SRP.
In addition, provide a list of those specifications required by ACI-359 that were not used for the Midland Plant.
130.15 Your terminology used in Section 3.8.6 to describe loads and (3.8.6) load combinations for steel and concrete structures is not consistent with that used in SRP 3.8.4.
Provide a cross reference for the two sets of tenninology. Our evaluation indicates that some of your load combinations may be less conservative than those delineated in SRP 3.8.4 Demonstrate that your criteria provide adequate margins of safety for plant design.
s.-
S e
.[~,
~
~
~_
..Z: _ '
~ J.
221-1 221.0 REACTOR ANALYSIS SECTION, ANALYSIS BRANCH 221.2' Provide the radial pressure gradient in the upper plenum and (4.4.2.5) at the core outlet for each allowable loop configuration.
Provide an explanation of how these effects are included in the thermal-hydraulic design calculation. Discuss, and support by calculations, the differences in hot channel pressure drop, flow, enthalpy rise and minimum DNBR relative to the assumption of a uniform pressure at the core boundaries.
221.3 The available experimental data for verification of the codes (4.4.5 and used for predicting the plant response to transient events are Chapter 14) limited. Provide the details of your proposed startup test program to obtain the data to verify the analytical methods and to demonstrate the transient characteristics of the plant. The program may reflect the tests performed at similar facilities which are applicable for verification of the Midland analysis.
221.4 Provide a description of the instrumentation available and the (4.4.4.5.6) surveillance requirements and procedures which would alert the reactor operator to an abnormal core flow or core pressure drop (e.g., due to crud buildup) during steady-state operation.
221.5 Discuss the basis and analytical procedure for establishing the (4.4.4.4.1) control band limits of -65 psi on primary coolant system pressure
'and +2*F on average temperature as those values maintained by the integrated control system (ICS). Justify the use of these limits for less than 4 - pump operation. Justify the assumption that the ICS maintains the system pressure and temperature within con-trol band limits.
Describe the load rationing features of the ICS which control the outlet temperature from each steam generator to prevent gross temperature maldistribution at the core inlet.
,r
~ - - -
i._.
232-1
- - - s 232.0 CORE PERFORMANCE BRANCH: PHYSICS SECTION 232.1 Several topical reports referenced in Section 4.3 have not yet (4.3) been submitted, provide these reports for review or provide detailed summaries of their content on the Midland docket. These reports include:
BAW-10119 BAW-10123 BAW-10121 BAW-10120 BAW-10116 BAW-10118 BAW-10122 232.2 The incore instrumentation is capable of detecting gross distortions (4.3.2) in radial power distributions. Hcwever, it may not be capable of detecting localized perturbations (e.g., interchange of Batch 1 and 2 assemblies near core center). Show in Section 4.3.2.2.7c that any fuel loading error that is not detectable with incore instrumentation produces perturbations which do not violate thermal limits when operating at 102% of full pcwer.
232.3 Have complete analyses been performed to identify all maloperations (15.0.1) or failures in the ICS or ICS control functions which might produce more serious consequents in transients?
232.4 Have safety-related systems also been analyzed with regard to (15.0.2) failure of passive components?
232.5 Discuss the consequences of the Startup Accident as a function (15.4.1) of initial core reactivity and indicate the reason for the choice of 1%ak/k subcritical.
232.6 Since scram occurs at powers significantly less than full. power (15.4.3) when a pressure trip occurs and the delay time of 0.7 seconds for the pressure trip is at the extreme of the sensitivity analysis, discuss:the suitability of.the full power scram insertion c'orve for the startup accident. What effects compensate for the fact that, at low power,.the fractional initial reactivity insertion is lower than that shown in Figure 15.0-37 232.7 Explain the statement (page 15.4-8) that the positive reactivity (15.4.3.1) increase due to single rod withdrawal will cause the inlet temperature to increase in view of the fact that the ICS acts to reduce inlet temperature with increasing power above 15". of full power.
4
.. ~
- __u
. m, c,
232-2
~
232.8 Explain the significance of the location of the point labelled (15.4.3)
"ICS compensation" on Figure 15.4-20.
In particular, indicate why it is plotted at an initial power of 105%.
232.9
. Describe the techniques used, assumptions made, and results obtained (15.4.8.4) which support your conclusion that no serious core damage or additional loss of coolant system integrity results from the rod ejection accident.
232.10 Describe the. features of the CRCM and reactor coolant boundary (15.4.8.2) design that prohibit or render very unlikely the ejection of a second control rod as a result of the first ejection.
232.11 The full power rod ejection accident is performed for an assumed (15.4.8.2) rod worth of 0.65%. What is the expected worth of a rod which might.:be ejected at full power? What scenario is employed to produce a rod of 0.65% worth? What would be the worth of a second rod which might be ejected?
~-
m 0
e
~
m 4
.urr 312-1
-4 312.0-5 CTIT! 3, ACC!3E iT Ent' ISIS 112"C3 T
312.29 The response to Request 312.8 is inadecuate. The referenced (3.5.1)
Gener:1 Electric mero r:pcet of " arch la, '073 limits the di:-
cussion of the tur ine overspeed pr taction systan tasting to its cacability of being tested during norral o;eration. A specific periodic testing progran is c'ai. ad by the repcet as existing in General Elactric "instructicn books."
Discuss briefly any plans for irplementing the valve tast pro-gram that is indicated in the General Elac.ric turbina genera-ter operating instructions and describe the program's salient features (e.g., valve tasting and inspacti:n procedures and frequencies).
312.30-The' results of your turbine missile analysis, as presented in (3.5.1)
Table 3.5-3, are in terms of the total probability Pa for damaging each of the Categ:ry I ecui; rent listed in tha table.
Please revise the taole so that the individual probabilities P2 and P3 are indicated s'eparately for each target.
312.31 It appears that the two biggest contributors to the overall (3.5.1) risk from low trajectory turbine missiles are the prinary system boundary within the containment and the safety related systa:s within the auxiliary building.
In reference to these areas, indicate if your striha and damage analysis includes the effects of cencrets scaboinc. Ces:rfba the modeling that is used in calculating the darace probability of safety related equipment due to concrete scabbing.
312.32 In reference to the response given to request 112.8, the cg (3.5.1)_
data presanted in Table 2.2-3 are not presentac in ter s of aircraft type. As indicated in Section 3.3.1.6 of the Standard Raview Plan, this information, along with affective plant area and crash probability per square mile are needed in order to estimate tne aircraft crash probability for tha site.
In order that we pay cceplete our evaluati:n, please cravide appropriate data and an analysis as cu-linad in Section 3.5.1.5 of the Standard Review Plan regarding th aircraft crash ha:ard 4
for the ::idland plant due to sir:raf t oparttf ans at the Carst:.!
and Tri-City airports. Uith res ect to e::h aircraft type, use aircraf t yearly operations projected f r th: life of the plant in your analysis.
312.33 Provide the fcilouing information:
(E.1.2)
A.
The type of paints and ccatings used in containment.
3.
The surface area and thickness or tot:1 r ss af each type
- of paint.
~
f me
-e 312-2 312.34 Provide tha type and amount of plastic and other or:anic (5.1.2) materials in contairment such as electrical insulati:n, machinery lubricants, caulking and seals, etc.
~
312.35 Ue are reviewing the c:ntrol rcom habitability systats with (5.4) respect to the chlorine ha:ards presented by the liquid
-(3A.l.95) chlorine storage facility at the Midland Cow Chemical sita and also by the nearby railway shipments of chlorine.
In accordance with Regulatory Guide 1.g5, adequate protection against chlorine for the Midland.*!uclear Plant would be provided by Type IV or Type VI centrol rooms.
The proposed control rocm ventilacion system does not ccmoly oith either of the above control reca categories for the follcwing reasons:
A.
The air exchange ratc during normal olant operation is greater than 1.0 air change per hour.
B.
The air exchange rate. for an isolated control rcce is estimated to exceed 0.015 air changes per hour under 1/8-inch water gauge pressure differential across all penetra-ttons.
w C.
If the control roem were to satisfy the Type VI criteria, control room isolation actuation by rem.ote chlorine detectors (i.e., detectors located at the potential point of chlorine release).iould be required.
Review your contrei r:ca habitacility systams design _in ref-erance to Regulatory Guide 1.95 and provide appropriata design modifications in order to achieve the level of protecticn m
against chlorine indicated by Table 1 of the guide.
Indicate those areas (if any) where your design will deviate signifi-cantly frem the reccceendations outlined in items C.1 thenugh C.G of the guide.
In particular, since the present and pro-facted quantities of chlorine stored at the Cow Chemical sita are large, long-term releases of chlorine may be pc:sible.
Thus, items C.4, C.5, and C.5 should be addressed in detail in-the FSAR.
4
'312.36 Discuss briefly the detection sensitivity of the toxic gas
((7.3))
detectors listed in Table 7.3-2, specifically in referance to 5.4 the chemicals identified in Table 5.a-3.
Includa in the dis-cussion a brief description of the toxic gas detectors in terms of basic principles of cperation.
'312 07 Provida data. showing the pri ary-to-sacendary leak rcte versus
' (15.t.C) time from accident initiati:n until the secondary and primary sy: tem pres ures ecuilibrata.
P
- ~..
'.6 313. ~,
312.03 The r1sponsa to request 312.24(a) is inadequate, Please provide
~ (15.4.5)'.
far the rod ejaction accidan. vith loss of offsita pov.ii curves showing the primary and secondary systers ter::erature and i
prassures versus time, for a period of two ".ours.
312.29 The rasocnsa to request 312.25(a) is inadequate. The revised (15.5.3)
Figure 15.5-3 does not provida the requested sec:ndary system temperature or pressure, or primary system pressure fcr tne two-hour period following the accident.
?lasse provide curves shcwing these parameters.
312.40 Sinca fuel handling operations for the tidi:nd plant are pro-(15.5.3) posad to take place while the containment is cnan to the environ-cent, we recuire that adequate measures exist to mitigate the consequances of a postulatad fuel handling accident inside containcent.
This can be accomplished either by prompt detection of any radicactiva release by use of radun-dant radiation detectors followed by aut:matic contain ent isolation, or by purging the c:ntain7ent via ESF grade filters.
The staff considers acceptable nitigation for this even; to be the critacion given in Standard Review Pian Section 15.7.4 that the dose should be well within the guideline values of e
10 CFR Part 100 (taken to be 255 or less of the Part 100 values).
Since oue preliminary evaluatien indicates the desa exceeds our acceptance criterion without or:cpt con.ainmen: isolation or other suitable mitication, provide a full response to the fiscs listed in our earlier request 312.28 and indicate hcw you plan to c:mply with this position.
312.41;)
The staff recuires, based upon Standard Reviaw Plan Section 15.5.5, (12.7.-
Appendix B that the dose consequences fron lea' age of ESF ecuip-ment such as pumps, seals, etc., following a desica basis LCCA be acceotably mitigated. Accepticle citignien may consist of release to the atmosphers af ter filtration through an ESF filter system, or by other suitable means such as an anpreoriatalv desi. nad leak-off collection systsa. Your FSAR does not indicate 'that such suitable mitigation of ESF leakage has been provided in the Midland
~ design. Revise your design acccrdingly and discuss your revised
~~
(Also, see rela ed request '21.5).
conformance with our position.
e
7 n
~
321-1
- - n 321.0 EFFLUE*1T TREATMENT SYSTEMS BRANCH 321.5 Your response to Acceptance Review Request 321.1 indicates that the (6.5.1)
Engineered Safety Features (ESF) Ventilation system designed to maintain (3A) a suitable environment for ESF equipment, and described in FSAR RSP Section 9.4.5 is not an ESF filter system and need not be designed
~
in accordance with the recommendation of Regulatory Guide 1.52 (Revision 1). However, it is our position that an ESF filter system is needed to control offsite doses resulting from pump leakage in post-LOCA operation. We require that you provide an ESF filter system as part of your ESF Ventilation System that satisfies all of the positions of Regulatory Guide 1.52 (Revision 1) for this purpose.
Revise your discussion in Appendix 3A accordingly.
O
.s v
.u w
u c.
7
~ =..
331-1
- < - s 331.0 RADIOLOGICAL ASSESSMENT BRANCH 331.2 Additional information is required regarding the measures you have (12.3) taken.to control buildup, transport, and deposition of the activated corrosion products in the reactor coolant and auxiliary systems.
In addition to your discussion on methods used to minimize piping low points, dead legs and crud traps, describe any steps you have taken to minimi:e'the buildup, transport and deposition of Co-58 and Co-60 in the reactor coolant and auxiliary systems. Examples of some methods used to reduce the formation and transport of crud products would include:
- 1. -The use of reduced nickel in primary coolant system alloys 2.
Low cobalt impurity specifications in primary coolant system alloys 3.
The minimization of high cobalt, hard facing wear materials in the primary coolant system 4.
The use of high flow rate /high temperature filtration.
331.3 Specify the frequency of calibration for your area and airborne (12.3.4) radioactivity monitors.
v-331.4 Provide a detailed layout of the solid radwaste area (12.3)
(similar to the one in Figure 11.4-5) indicating radiation :oning for the 634-foot and 652-foot levels.
331.5 Indicate whether local exhaust systems will be installed in the (12.5.2) hot machine shop for work on contaminated items.
If such systems are not planned, identify the alternate measures planned to limit airborne contamination from the machining of contaminated items.
4 e
-.e.~.
~
362-1 362.0
.'GEOTEC11NICAL' ENGINEERING 362.1 Provide a summary of the results of field density tests for (2.5.4.5.3) compaction and moisture control of structural fill beneath and adjacent to Category I struc*ures.
362.2 Question 1 and the resulting discussion on page 8.00-1 included.in (2. 5. 4. 5.1)
Amendment Number 9 to vour PSAR stated that all natural sands with.
relative densities less than 75% would be remved beneath all Class I structures and beneath non-Class I structures so sited that their failure could er. danger the adjacent Class I structures. Discuss the methods ' employed in mapping and re=oving the sands having less than 75T relative density. Provide plan and sectional figures p
showing the areas where these materials were removed. Figure A9-2 of the PSAR which displays sub-s' rface profiles of Class I piping u
should be updated to show removal of sands of less than 75% relative density and be presented in the FSAR. Figure 2.5-21 of the FSAR shows loose sands beneath ti e Class I tanks although they were to have been removed. Explain'this inconsistency, and provide proper documentation of as-built conditions.
362.3 Reference is t,ade in section 2.5.4.10.2.3 to Table 2.5-14 for design (2. 5. 4.10. 2. 3)
. values of passive pressure. The table number is incorrect and should read Table 2.5-15.
e 9
e y
y v.
e--
A n
w 362-2
.e 362.4 Provide the results of all benchmark survey measurements taken (2.5.4.13) during construction. Graphically, compare the measured results to predicted settlements. Provide a commitment and schedule to
. submit ' che results of future survey settlement measurements.
362.5 Provide gradation curves for the 12 inch thick crushed rock bedding (2.5.6.4.2) layer beneath the riprap. Discuss the adequacy of the bedding material with respect to the requirements of a filter.
362.6 Provide figures showing the f ailure surfaces that resulted in the (2.5.6.5.3) minimum cocputed factors of saf ety for all slope stability conditions studied.
362.7 Paragraph four of section 2.5.6.5.4 states that the outer slope of "s-(2.5.6.5.4) cross-section I was used to simulate the plant area fi31 and a seismic coefficient of.12g was used. Howevar, Table 2.5-20 indicates that cross section G was used for this condition. Explain and correct this inconsistency.
362.8 Provide a detail of a typical pie:ometer as installed in the (2.5.6.8) cooling-cond dike. Also provide cross sections showing the development of<the phreatic surface from initial piezemetric head to full pond steady-state condition and a comparison to the phreatic surface
- assumed for the stability analysis of the steady-state cendition.
l 1
i
.-q
~
O m
371-1
~
371.0 HYDROLOGIC ENGINEERING 371.9 You state that areas adjacent to power plant structures and site (2.4) drainage facilities are designed for a rainfall intensity of 6.1
-(RSP) inches per hour which corresponds to a 100-year precipitation event.
In previous paragraphs, you state that the plant drainage is designed for the 24-hour Probable Maximum Precipitation (PMP) of 13.0 inches. You apparently conclude from this that the 100-year pre-cipitation.is more intense than the PMP, since 6.1 in./hr is more intense than 13.0 inches in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This approach is incorrect.
The 24-hour PMP must be broken down to appropria*.e time increments, suitable for the drainage areas and times of concentration which exist at the site.
It is our position that site drainage facilities (with times of concentration of about 10 minutes) be designed for the local PMP rainfall intensity of 20 in./hr. This intensity corresponds to a longer-period Pf1P, broken dcwn into appropriate time increments for small drainage areas. Document the adequacy of your design by providing an adequate response to Request 371.1 using the rainfall intensity described above.
. s.
r ese i
e d
w
b s
m n
372-1 S72.0 METEOROLOGY 372.12 The_ design basis temperatures used for the design of the Midland (2.3.1) plant heating, ventilating, and air conditioning systems are given as 96F dry bulb and 79F wet bulb for su=mer and -10F dry bulb for winter. For what duration of time would these temperature values have to be equalled or exceeded before operation of the heating, ventilatins, and air conditioning systems would be af fected?-
372.13 The:onsite stability distribution for Midland.(3/75-2/7/) was (2.3.2) compared to the stability distribution frem several other sites, (Greenwood (9/72-9/73), Erie _(11/73-10/75), Cook (5/75-4/76),
Quanicassee (6/72-5/73)]'
While Midland showed 57% D stability and 21% E stability, the other sites ranged from 25% to 38% for D stability and 33% to 47% for E stability. Although year to year variability may lead to a biased distribution,.the 57: D stability found at Midland is based on two years of data.. This lessens the possibility that the large amount of D stability is based on an anc=alous year. Although the other four sites cover a variety of meteorological regi=es, they all show similar stability distributions (including Quanicassee which is based on eg).
l
'dowever, Midland does not show good agreesent with these other sites. Discuss further.the validity of the onsite stability.
j '
i distribution based on the onsite data at the Midland Plant. In particular, look at the method that is used to determine vertical temperature differences.
372.14~
Provide monthly joint frequency distributions based'en the onsite (2.3.2)-
data from the Midland plant for the 60-10 eter vertical te=perature difference and 10 meter winds for the period March 1975 through February 1977.
372.15 Provide yearly joint frequency distributions based on the Flint (2.3.2) data and STAR stability classification for the years 1972 through 1977.
372.16 Are temperature difference (ATs) =easured directly or are they (2.3.3) determined by subtraction of the temperature at two different levels?
e -
372.17 fin the event an-instrument outage renders meteorological data (2.3.3) unreceivable.by the teletype in the plant control room, "such data will be available via telephone". Where will the telephone supplied meteorological information come from and of what will it consist?
4 8
a 4
4 1-
-400 1
-s 400.0 PROJECT MANAGEMENT 400.1 FSAR 5ection 1.6 indicates that you place relf ance for the safe (1.6) design of the plant upon several topical reports which have not (RSP) yet been submitted for our review. We require, prior to' issuing the SER for Midland Plant Units 1&2, that all such reports be sub-mitted, our review completed, and changes for the. Midland docket made as may be appropriate based upon our final approved version of the report. ~ Accordingly, it is in your interest to expedite submittal of these reports consistent with your review cchedule so that they can be scheduled consistently. Your intended siQmittal date for each such report should be indicated after your statement "to be submitted." Also, you should check Section 1.6 for completeness since several topicals referenced in subse-quent text sections have been omitted (e.g., see request 232.1).
9 4
_ +
w v
. ~.
I 421 -1
.421.0 QUALITY ASSCRANCE 421.1 Identify or reference those safety-rolated structures, systems, and components under the control of the Midland QA program.
421.2 In Topical Report CPC-1-A, Rev. 5 you connit to conply with WASH documents with certain acceptable alternatives.
Since the WASH documents contain a numoer of draft standards which have been superseded and since the Midland FSAR was docketed on 11/18/77, there are Regulatory Guides and Regulatory Guide revisiona issued prior to this docket date which apply and should be addressed in the Midland FSAR.
These include:
RG 1.33 Rev. 1 RG 1.94, Rev.1 RG 1.116 Rev. 0-R RG 1.38 Rev. 2 1.123^Rev. 1 Accordingly, revise your Report CPC-1-A, Rev. 5 to delete the comitment to the WASH documents and to provide a specific cemitment to comply with the Regulatory Position of the following Regulatory Guides and the requirements of the following ANSI Standard:
RG 1.8, Rev. 1-R RG 1.28, 6-7-72 RG 1.30, S-11-72 RG 1.33, Rev. 1 RG 1.37, 3-16-73 RG 1.38 Rev. 2 RG 1.39, Rev. 2 RG 1.58, 8-73 RG 1.64, Rev. 2 RG 1.74, 2-74 RG 1.88, Rev. 2 RG 1.94, Rev. 1 BG 1.116, Rev. 0-R RG 1.123, Rev. 1 ANSI N45.2.12, Draf t 3, Rev. 4, 2-74 Any exception ~s, alternatives, or clarifications you believe warranted should be identified with sufficient supporting detail.
Your FSAR should then reference this revised topical by specific revision number and should include any plant specific exceptions or alternatives you consider appropriate.
6 9
~.
w.
, mm.e -
Sen
-: - a 422-1 422.0 CONOUCT OF GPERATIONS 422.3 Describe tne responsibilities and authority of your Staff (13.1.2.2) health Physicist, and Staf f Chemist snown in Pigure 13.1-3.
422.4 Describe your qualification requirements for the positions (13.1. 3.1 )
of Staff Health Physicist, Staff Chemist, Quality Control Supervisor, and Electrical Supervisor.
' 422.5 We do-not agree with, or need clarification regarding the (13.1. 3.1 )
qualification requirements described in Section 13.1.3.1 for RSP several positions. Below is our recuest for clarificaticn or a statement of our staff position, relative to these
- osi tions
J 1.
Operations Superintendent - It is our position that tne Operations Superintendent should hola a senior operator's
' license, whether or not the Plant Superintendent has a t
senior operator's license.
2.
Health Physicist - It is our position that three of the five years experience be es stated in Revision I to Regulatory Guide 1.3; i.e., "be applied raciation protection work in a nuclear facili ty...."
3.
Chemical Engineer - This should be clarified to assure tna; the 1 year experience is in radiochemistry.
4.
Maintenance Repairman "A"/ Maintenance Electrician "A"/ Senior Technician - Please clarify such that tnese ' positions are comparable to tnose described in Sections 4.5.2 and 4.5.3 of ANSI N18.1-1971; i.e., technicians and repairmen in responsible positions.
422.6 Describe the specific position title of each memoer of your (13.4. 3. 3 )
Safety and Audit Revis.. Board (SARB) or describe tne qualifica-tion requirement for members of the SAR3.
e
-w w-m y
r-,
=
- - _.?
._2_.
L..
_ _.=
~
,~
t
. _. 441 -1 441.0 OPERATOR LICENSING 3 RANCH: TRAINING SECTICN 441.1 Amend section 13.2.1.2, to include a cannitment that refresher (13.2.1) training for cen-licensed persennel shall be periodic and con-ducted at least every two years. This training shall include,
. at a minimum, refresher training en administrative, radiaticn protection,emergencv and security recedures.
441.2 Amend section 13.2.1.5.1, to include a : mmitment that periedic (13.2.1) written qui::es will be administered at the conclusion of each lecture or series of lectures.
all.3 Change Iten (e) in Section 13.2.1.5.3.1 f rom the present werd-(13.2.1) ing to, "Any sicnificant ( 10f.) power change in manual red con trol." Recove the statement that refers to these reactivity
==
changes as examoles and crevide a :amnitment that these listec reactivity changes are these ac:eptable to meet the ccerator
-c.
recualification crcgram.
041.4 Anend Section 13.2.5.5, to pecvide a cccmitment that not more (13.2.1, tnan two 'vcensed individuals who administer the examintiten shall be exemot fr:n the annua! written examingtf on.
441.5
.In secti'cn 13.2.1.5.7, it is stated that the c:erational eval-( 13. 2.1) untion can be comoleted at a simulater. This evaluatien at the simulator is only acceptable if the re:uirenents of 10 CFR 55 Appendix A Item (3) (d) are met..If these requirements cannot be met, provide a cc7nitment that the ocerational evaluation will also include evaluation of performance at the plants 441.6 Anend Section 13.2.1.5.8, to include a conmitment that recceds (13.2.1) shall be maintained to include cocies of written examinations, the answers given by tne licensee, results of evaluations, and documentatien of any additional training administered.
A41.7 Amend 13.2.1.5.2, to include a c:cmitment that all licensed in-( 13.2.1) dividuals shall be cecnizant of facility design changes, pro-
.cedure changes and facility license char.ges and that auditable.
records are maintained.
w*
M..d p
_w--
Weeks m e + go.
'**h9
. g s
%_6
s.
-m
~
w 441-2
' ~ ~ #
442,0 OPERATOR LICENSING BPANCH:
PROCEDURES SECTION 442.1 In section 13.5.3, include a cemitment that the administrative (13.5) procedures shall include the requirements to meet 10 CFR Part 50.54 (i),(j),(k),(e),and--(m).
442.2 Provide a comitment that the administrative and operating pro-
~
- (13.3) cedures shall be completed six months prior to fuel loading and revise Figure 13.5-1 to reflect this comit:nent.
.*s*
3 m
J
--tha m
O e
p y
7
b
+
6^- s,;
-i.-
7_ _
o e 2
m
_?
,,)
REQUEST : FOR ADDITIONAL INFOR.'!ATION yin, wn Distribution:
NRC-PDR.
LocaLPDR Qlocke t FITE LWR *4 File R. S. Boyd-R. C. DeYoung
.D.
B. Vassallo F. J. Williams -
S. A. Varga Proj ect blanager DARL H000 il St.' Service R. J.-h!attson D. Ross J. Knight R. Tedesco l
H. Denton V. A. >!oore i
R. H. Vollmer
- 31. L. Ernst W. P. Gammill-W. FicDonald l
~ ELD IE (3)
W. Haass i
bec: ACRS (16)
T. Abernathy_
J. R. Buchanan I
4-
. j i
4 w,
+ +, -
,n.
y t-n e-9 y,e
.-ea+
-,-e e,
ea w
e r~--
5
-~-