ML19289C684

Forwards Response to Request for Addl Info on QA & Subcompartment Analysis.Info Will Be Incorporated Into FSAR in an Amend to Application

ML19289C684
Person / Time
Site:	Salem
Issue date:	01/18/1979
From:	Mittl R Public Service Enterprise Group
To:	Parr O Office of Nuclear Reactor Regulation
Shared Package
ML18078A672	List: ML18078A671 ML18078A673 ML19289C684
References
NUDOCS 7901220101
Download: ML19289C684 (33)

Text

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O PSEG Publ.c Service Electnc and Gas Company 80 Par, P,;m fe..a'* N J 07 0 ' P* - . U e :',7'//;

January 18, 1979 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Mr. Olan D. Parr, Chief Light Water Reactors Branch 3 Division of Project Management Gentlemen:

RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION NO. 2 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-311 Public Service Electric and Gas Company hereby transmits sixty (60) copies of its responses to certain of your requests for additional information to NRC Questions 4.38, 5.96, 5.110, 13.9 (preoperational testing), Quality Assurance and Sub-compartment Analysis. The information contained herein will be incorporated into the Salem FSAR in an amendment to our application.

Should you have any questions, please do not hesitate to contact us.

Very truly yours,

./ .Y J

./ '

R. L. Mittl General Manager -

Licensing and Environment Engineering and Construction Enclosure 790122010\

Thr- Enera Peopio

s QUESTION 4.38 Appendix XVII-2461 of the ASME Code Section III requires that bolt loads in bolted connections for linear component supports include prying effects due to the flexibility of the connection.

(1) Provide confirmation that the loads in bolted connections for linear compenent supports were determined by consider-ing the deformation of the cor,ection and tension-shear interaction for the bolts. For connections of supports which are anchored to a concrete structures provide in addition:

a. The type of anchor bolt

b. The factors of safety (and their bases) against pullout under static, repeated and transient loading.

This information should include representatives diagrams of the connections, material properties and interaction diagrams, the analythical techniques and models used, and the maximum stresses in the bolts and the connections under both static, repeated, and transient type loading.

(2) If any connection was assumed to be rigid, provide complete analytical or experimental justification for this assumption.

AUSWER (1) In developing designs for bolted component supports for piping, tension and shear interactions were considered.

The design conservatism on structural members is con-sidered sufficient such that deformation of the connection does not adversely affect the capacity of connections to withstand design loadings.

a. Types of anchor bolts used for the various bolted connections in the plant are as follows:

(1) The majority of safety related supports employ connections bolted to concrete inserts which derive their strength from an integral steel

coil embedded in the concrete at the time of structure forming.

(2) The other type of anchor bolt used employs an expandable vedge piece inserted in a pre-bored hole in the concrete.

b. Loads applied to these anchor bolts are within manu-facturer's specified limits. Attached you will find representative analysis of typical standard supports.

(2) The assumption of rigidity for bolted linear support con-nections, where applicable, is made on the basis that the applied loads to the supports have been determined by analytical methods to be adequate. Refer to attached typical support evaluation.

. ATTACHMENT QUESTION 4.38 RESPONSE

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b gg = x (0.& O(2)(/ 4 f) = o. 73 k 5

45

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% TAL Ed . /u/ fod ONE G WP = g + A f = 0. Pf o.9

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(Pu>s 236 8 5 89 4253)

Conswep loAp = 474T/45

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2 - EX AnPLE (LCAO ACTIMG AT hipSPAN 1L=8; B= = 2.62 4 ; D = f = /.54 ; As =0.61 L' (i + setz L = +3 4 ; 4 =4.o4) Z = 916 4! ~ 2 - C 4x 7 25 s n e.q '+ u,.g s ec.so '

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=

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2 2e c.5 /

TOTAL R.EACTICM FCC CNE EMD -* R+b R = C 50f + fr4l 7 = /*9l ?

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A ssunPTICAIS FOR ccNcRETE ANCH0AAGE BEHAVICVR SA!7E AS fcR wcRKING STRESJ NET / fob, (1)

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PROOF TEST REPORT Pr od u c t l ' d t a:e te r P.i c h= nd E .C . Ty p e Insert with ce: hine thread coil pulled frc: IS" x IE" x 6' con: rete slat by meane cf l' x }6" Anchor Stud Bolt with n.ats. The insert was made of .442i. wire, and its setback in the concrete was 1/5 ' . The c o n: r e te s l at w a e r e i r."c r e e d w i'l a .442 wir e nat, 6* x 6' center eper.ing, lo:sted at mid-deptA ef the slat. Tr.e strength of the c oncre te wa s 25 j0 p . s.1., ar.d the slip dial indicator was zerced in at a lead of 2000 lbs.

Failure cceurred in both specimene by the insert pulling out of the concrete slat. Six cra:ks e anated frc: the incert er. the top cf the slab and extended down on four side surf aces te the reir. force:ent. The first crack appeared with a lead of about 14000 lbs. or. both specimens.

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A.:h:r S*wd Insert S;eciser N:. 1 Spe:1:en No. 2 Lead, ki:e Slir, in. Loed, kita Slic, in.

2 0 2 0 A O.021 4 0.005 6 C.026 6 0.C15 6 0 . 0.'.5 6 c.02) 10 0.0 f4 10 0.0L4 12 C .06 0 12 0.058 14 0.092 14 0.072 16 0.111 16 0.089 12 0 .1 62 IS 0.107 20 , 0.162 20 0.127 22 0.205 21 0.149 24 0.2A5 21- 0.153 Ultimate Loed : 25500 lbs. Ul tima te lead = 24630 lb s .

OUESTION 5.96 Provide the criteria used for the selection of the number of lumped masses.

ANSWER Refer to the response to Question 5.25. The containment structures at the Salem station used a finite element model for the seismic analysis. A total of 190 elements were used in discretizing the structure.

The Auxiliary Building and Fuel Handling Building used the lumped mass models for the seismic analysis. The points of mass concentration of these buildings are most apparently at the roof, floor and foundations. Heavy equipment and sub-systems in the buildings are rigidly attached to the floors.

Therefo  :, the masses of the analytical models were logically lumped at these levels.

We have reviewed our analytical models and have concluded that the degrees of freedom used are adequate. Additional number of degree of freedom in these models will not result in more than 10% increase in structural responses.

Based on the above, the Salem design is in compliance with the modeling criteria defined in Section 3.7.e of the SRP.

QUESTION 5.110 In request for additional information 5.100 we asked that you state if the fundamental frequencies of the key subsystems are controlled to be either greater than twice or less than one-half the dominant frequencies of their supporting system. Your response stated that the fundamental frequencies of the key subsystems were considered in relation to the dominant frequen-cies of their supporting systems. However, you did not state if the above criteria were used to accompli-h the adequate design of the key subsystems or some other criteria that ma' be proven to be just as adequate. Provide a more detailed response to this concern.

ANSWER The fundamental frequencies of key subsystems were considered in relation to the dominant frequencies o~ their supporting systems. Elimination of resonance was one of the principles of design. Various methods for seismic qualification were employed for key subsystems. In most cases, the key subsystems were considered to be very flexible and were analyzed / tested as a decoupled system from the supporting system. Refer also to the response to Question 5.38 which addresses the approach to avoid the predominant input frequencies of components to earthquake inputs. Refer also to the responses to Questions 4.12, 5.35, 5.37, 7.18 and 7.29.

These subsystems were analyzed / tested as a decoupled system from from the supporting system, because the mass ratio of the subsystem to that of the supporting system is less than 1%.

M,

. QUESTION 13. 9 (a ) (Initial Testing)

The test methods and plant's electrical systems status need to be defined for tests that will be conducted to satisfy regulatory positions in Regulatory Guide 1.41, "Preoperational Testing of Redundant Onsite Electrical Power Systems to -

Verify Proper Load Group Assignments." If exceptions to this guide are taken, they should be explained in sufficient detail

.to show that the plant status and test methods will provide equivalent assurance of proper load group assignments and independence between redundant AC and DC sources of onsite power and independence from offsite power sources. Otherwise, the test methods described in Regulatory Guide 1.41 will be required by the staff.

ANSWER The No. 2 Unit initial preoperational test program is in full conformance with the Regulatory guide which functionally demonstrates the independence among redundant onsite power sources and their load groups. This is accomplished by the performance of the Integrated Safeguards Test. As stipulated in part c.1 of the guide, isolation from the offsite trans-mission network will be accomplished by the direct actuation of the undervoltage sensing relays (opening the 4 kv AC undervoltage relay knife switches). All loads off the No. 2 Unit group buses not required to maintain necessary and independent construction and testing activities, as well as backup power to No. 1 Unit,will be de-energized to the maximum extent practical.

The functional testing requirements covered under c.2 and c.3 of this guide are performed as part of the Integrated Safeguards Test.

P78 170 48 Q13. 9 (a )

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QUESTION 13. 9 (b)

Our position relative to your proposal to eliminate the turbine trip test from 100 percent power is that it is not acceptable. PSE&G should be modified to include this test or the following additional information should be provided:

1. Provide a listing of all initiating events or conditions that result in opening of main generator output breaker.

2. Describe which trips listed in item I will result in a direct turbine trip event and which trips will result in a turbine trip event via sensed T-G overspeed conditions.

3. Describe the automatic transfer functions for the plants electrical distribution system along with associated time delays for each initiating event or condition.

4. Describe the means by which'you plan to initiate the generator load rejection test from 100 percent power.

ANSWER

1. Table 013.9-1 indicates the initiating events which cause opening of the main generator output breakers and those which result in a turbine trip. The initiating events are crouped to indicate whether they act to cause a direct turbine trip or a turbine trip via a common tripping device. Refer to Figure 013.9-1 for information regarding the breaker setup.

2. As indicated in Ta ble 013. 9-1, all automatic generator breaker trips (both breakers) will cause a direct turbine trip. Only a manual trip of both breakers will not cause a direct turbine trip. This will not cause a turbine overspeed trip but result in a reactor trip through primary system parameters and a subsequent turbine trip. An individual output breaker trip will not cause a turbine trip or overspeed P78 170 50 013. 9 (b) -1

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since the generator still has output to the electrical system via the remaining output breaker.

Operation of the unit with only one generator breaker in service is considered to be an abnormal operating condition used only during periods of maintenance which require the condition. Any trip signal which normally opens bo9h 7enerator breakers and causes a direct turbine trip is unaffected by having one breaker open prior to the trip signal. The trip signals listed in Table Q 13 9(b) which normally open only one generator breaker and do not produce a direct turbine trip would perform in the same manner if a generator breaker were open prior to the trip signal.

In this case, the unit would respond as if a manual trip of both generator breakers had occurred.

3 The automatic transfer function associated with these events is the transfer of the h kV group busses from the auxiliary power transformers to the station power transformers. This transfer is accomplished in less than one second.

Automatic transfer of the kV group busses is accomplished when all of the following conditions exist:

a. Both 500 kV generator breakers are open.

b. Potential exists on the Station Power Transformer.

c. None of the protective trips have occurred (i.e., bus differential, bus overload, and failure of auxiliary power transformer side infeed break'er).

4. The Generator Trip Test will be performed at 100% of rated thermal power. It is expected that a reactor trip and a turbine trip, as well as any turbine overspeed, will be noted and recorded. The test is performed by manually opening both output breakers.

P78 170 51 Q13 9(b)-2

Salen intends to conduct it itenerator trip test cJnce it will cause a more cevere trancient on the plant than rt turbine t ri p.

This van done on Unit 1 and i:: our .i n t e n t on Unjt 2.

A turbine trip vill caune an i.rmediate reactor t. rip when atove the p7 setpoint power level and vice veran, a r e rt e t o r trip vill alwayc cauce a turbine trip. Cince these two eventc occur in conjunction with each other, the differcnces in effee.c on the plant whether a turbine tripc firct or the reactor ic negligible. Since the data that vould be genclated from an additional turbine trip would be inciCnificantly different than that generated durinC a reactor trip and that the generator trip as deceribed abcre is the more severe trancient, the coctc accociated with e turbine trip tect do not appear to juctify the benefits to be derived from the tect. If during the cource of power operation prior to the first refueling an event such as thic does not occur, then a reactor trip (with cubsequent turbine trip) test will be performed.

p78 170 52 Q13 9(b )- 3

/_

OUESTION 13.9 (c)

The staff concluded that Regulatory Guide 1.108, " Periodic Testing of Diesel Generator Units as Onsite Electric Power Systems at Nuclear Power Plants" is applicable for the

. Salem 2 facility. Since this guide addresses both pre- ,

operational and periodic testing, PSE&G needs to be modified to describe how your planned preoperational tests will conform

- with this guide of how they will provide for equivalent pre-operational testing.

ANSWER The No. 2 Unit initial preoperational test program is in conform-ance with the Regulatory Guide 1.108 with the following exceptions:

Paragraph c2.a (4) - We comply with the section by tripping the diesel output breaker at 2750 KW (2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating) and verifying that the voltage regulation and overspeed limits are not exceeded. We feel this transient is more severe than the load shedding requirements identified in the regulatory guide.

Paragraph c.2a (5) - We will perform the test described in this section but due to the sequence of testing it may not be immediately af ter the test described in c.2.a (3) . The generator systems will, however, be at full load temperatures.

Paragraph c.2.a (6) - Our plant is not designed to perform the test described in this section.

O Ol3. 9 (c)-1

- Paragraph c.2.a(9) - To accomplish this reliability demonstration, we will increase the frequency of surveillance. testing.to. acquire the 23 starts per diesel prior to proceeding beyond the Zcro

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Power Physics Test Program. To accomplish this we intend to take credit for those diesel starts accomplished to date or scheduled during Integrated Safeguards Testing, as long as the diesels are loaded to a minimum of apprcximately 25% and the run durations are approximately 30 minutes or more. All additional starts will comply with the regulatory guide criteria for valid tests. ,

P78170 47 Q13. 9 (c) -2

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QUESTION 13 9(d)

The staff has concluded that Reg. Guide 1.68.2 " Initial Start-up Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants" is applicalbe for the Salen 2 facility. NRC requires that your application be corrected and modified to describe the tests planned to confor=

with the guide or show that equivalent testing vill be con-ducted. State your intent to comply with our requirement.

AUSWER The Salem plant was designed for remote hot shutdown fro outside the control room. This was described in Section 7.7 of the FSAR. Our capability to go to a cold shutdown condition through the use of procedures and te=porary modifications was dercribed in the Response to Question Q9.45.

General Design Criterion 19 of 10CFR50 Appendix A requires a design capability for remote hot shutdown with a potential capability for subsequent cold shutdown through suitable pro-cedures. A detail procedure vill be written explaining the actions to be taken to bring the plant fro = a hot shut-down to a cold shut-down condition from outside the control roon This procedure vill be completed by February 19, 1979 A trial run, not an actual test however, vill be performed to demonstrate the co==unication, coordination and capabilities of the procedure to achieve cold shutdown conditions.

A functional outline vill be submitted to the staff which vill present the high lights of the detail procedure

indicating how cold shutdown will be achieved. This will be transmitted for the staffs review by Fe'bruary 5, 1979

D.5.10 (CONT) p p 1 *

~ Inspections shall be performed by personnel who are qualified in accordsnee

\ .

with Regulatory Guide 1.58 as noted in D.5.2 Item g othe r than those who per formed the activity being inspected. When the inspection requires special skills (such as radiography), arrangements shall be made to have appropriate inspectors perform this work. These inspectors may be from within the company or from outside organizations.

Testing, repair and maintenance activities shall be inspected by qualified individuals other than those who performed or directly supervised the activity being inspected. Inspection of operating activities (work functions associated with the normal operation of the plant, routine maintenance, and certain techni.al services routinely assigned to the onsite operating organization) may be conducted by second-Lun supervisory personnel or other qualified personnel not assigined first-line supervisory responsibility for conduct of the wor.k. The signature of the respective supervisor on the work package signifies that all inspection and/or test require-ments have been satisfactorily completed or completed with non-co-formances, as noted. L Station Department Heads are responsible for inserting mandatory inspection hold points in procedures they approve. The Station Operations Review Committee (SORC) may recommend to the Station Manager, additional or different hold points, as a result of their review.

The Station QA Engineer can also require that additional inspection hold points be added to a procedure.

The Station Administrative Procedure Manual establishes the require-ment that Station Department Heads are responsible for the prepara-tion of procedures for activities affecting nuclear safety and for the SQAE and SORC review of such procedures prior to implementation.

Department manuals identify the requirement that hold point inspections must be considered for inclusion in procedures.

When a " Hold for QA Inspection" is identified, a qualified individual assigned by the SQAE will perform the inspection.

D 5-21