ML20148C378

Provides Updated Final Listing of Issues Originally Addressed in Category E of 771212 NRC Qualification Review Ltr.Util Should Respond to Issues & NRC Requests for Addl Info

ML20148C378
Person / Time
Site:	Palo Verde
Issue date:	10/12/1978
From:	Parr O Office of Nuclear Reactor Regulation
To:	Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
References
NUDOCS 7811020030
Download: ML20148C378 (11)

Text

- - -_ - _-__ _ __

h] (L L PDL

1. % , UNITED STATES 3'

f d NUCLEAR REGULATORY COMMISSION j.I'D j WASHINGTON, D. C. 20555 OCT 121978 Docket Nos. STN 50-592 and STN 50-593 Mr. Edwin E. Van Brunt, Jr.

Vice President, Construction Projects Arizona Public Service Company P. O. Box 21666 Phoenix, Arizona 85036

SUBJECT:

RADIOLOGICAL SAFETY ISSUES TO BE REVIEWED FOR PALO VERDE NUCLEAR GENERATING STATION, UNITS 4 AND 5 In our qualification review letter dated December 12, 1977, we listed several significant safety issues that had been identified since the issuance of the Safety Evaluation Report for Palo Verde Units 1, 2, and 3. These matters were listed in Category E of the qualification review results. We also stated in the December 12 letter that it is our intent to request that you address any new staff positions which are developed during the Palo Verde ,

4 and 5 review and which satisfy the replication policy. The subject of staff positions which are to be applied to Palo Verde has been discussed in our meetings of August 4 and September 28, 1978. We informed you that June 30,1978 had been established as the cutoff date for new staff positions.

l The purpose of this letter is to provide an updated, final listing of the l

issues originally addressed in Category E of the qualification review l letter. Enclosure 1 lists nine issues which replace items E.1 through E.7 '

of the qualification review letter. You have already been requested to address these matters either via the qualification review or by staff requests for additional information. Enclosure 2 lists 16 issues which replace items You hrve been requested E.8 through E.26 of the qualification review letter.

to address these matters in the qualification review with the exception of Regulatory Guide 1.141. You should describe how the Palo Verde design meets the recommendations of Regulatory Guide 1.141. Finally, Enclosure 3 lists matters which replace items E.27 through E. 34 of the , qualification review letter. Of the additional items added in this category since the qualification review, the staff will determine those for which we need edditional information and will provide requests to you.

7 L _ ___ _ ___ _ _ _ _ _ _ _

OCT 12197B

., e.

Should you have any questions or require clarification of these matters, please contact us.

Sincerely, Olan D. Par'r, YIiief b

Light Water Reactors Branch No. 3 Division of Project Management

Enclosures:

As stated ces w/ enclosures:

See next page 1

l l

I

1. 4

Mr. E. E. Van Brunt , Jr. cc: Arthur C. Gehr, Esq.

Snell & Wilmer 3100 Valley Center Phoenix, Arizona 85073 Charles S. Pierson, Assistant Attorney General 200 State Capitol 1700 West Washington Phoenix, Arizona 85007 Donald G. Gilbert, Executive Director Arizona Atomic Energy Commission 2929 Indian School Road Phoenix, Arizona 85017 George Campbell, Chairman Maricopa County Board of Supervisors 111 South Third Avenue Phoenix, Arizona 85003

" m

1. 9 6

• ENCLOSURE l_

DOCUMENT NUMBER REVISION TITLE

-= -

RG 1.99 1 Effects of Residual Elements on Predicted Radiation Damage to Reactor vessel Materials (Paragraphs C.1, C.2 and C.4)

RG 1.101 1 Emergency Planning for Nuclear Power Plants

. RG 1.114 1 Guidance on Being Operator at the Controls

- of a Nuclear Power Plant RG 1.121 0 Bases for Plugging Degraded PWR Steam

. Generator Tubes

, RG 1.127 1 Inspectior af Water-control structures Associated with Nuclear Power Plants

• RSB 5-1 1 Branch Technical Position:

Residual Heat Renoval System RSB 5-2 0 Branch Technical Position:

Reactor Coolant System Overpressurization Protection (copy attached)

RG 1.97 1 Instrumentation for Light Water Reactor Nuclear Power Plants to Assess Plant Conditions During and Following an Accident (Paragraph C.3 - with additional guidance on paragraoh C.3.d to be provided later)

RG 1.68.2 1 Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants O h

- - - ~ _ _ _ _ _ _ _ _ .___m_____

l

- - - ENCLOSURE 1 CRAllCH TECHillCAL " '.,i flCil R50 5-?

OVERPRE55URlZAT10fl PROTECT 10fl 0F PRESSURlZED WMER REACTORS.

• WHILE OPERATillG AT LOW TEMPERATURES

, A. Backqround General Design Criterion 15 of Arpendix A,10 CFR 50, requires that "the Reactor Coolant System and associated auxiliary, control, and protection systems shall be designed with suf ficient margin to assure that the 1

desion conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences."

Anticipated operational occurrences, as defined in Aroendix A of 10 CFR 50, are "tho(c conditions of normal operation which are expected to occur one

• cr mnre times during the life of the nuclear power unit and include but are not limited to loss of power to all recirculatien pumps, tripping cf the f.urbine generator set, isolation of the main condenser, and loss of i

all offsite power."

Anpendix G of 10 CFR 50 provides the fracture toughness reauirements for reactor pressure vessels under all conditions. To assure that the

[* Appendit G limits of the reactor coolant oressure boundary are not exceeded during any anticipated operational occurrences, Technical Specification pressure-temperature limits are provided for operating the plant.

The primary concern of this position is that during startup aad shutdown conditions at low temperature, especially in a water-solid condition, the reactor coolant sy', tem pressure might exceed the reactor vessel pronoro-temperature limitations in the Technical Specifications established for protection against brittle fracture. This inadvertent evorpressurization coJld be generated by any one of a variety of mal-functions or operttor errors. Many inciderits have occurred in operating plants as described in Reference 1.

Addit.ii:nal discussion on the background of this position is contained in Reference 1.

t , '.

e 9

t ENCL.05URE 1 _

B. Branch rosition c

1. A system should be designed and installed which will prevent exceeding the applicable Technical Srecifications and /ppendix G limits for the reactor coolant system wnile operation at low temperatures. The system should be capable of relieving pressure during all anticipated overpressurization events at a rate suf ficient to satisfy the Technical Specification limits, particularly while the reactor coolant system is in a water-solid condition.

2. The system must be able to perform its function assuming any sinole active component failure. Analyses using appropriate calculational techniques must be provided which demonstrate that the system will provide the recluired pressure relief capacity assumina the most limiting single active failure. The cause for initiatien of the evnnt, e.g. , operator error, component malfunction, will not be consi'lered as the single active f ailure. The analysis should assume the most limiting allnuable operatinq conditions and systems configuration at the time 'of the postulateo cause of the overnressure event. all notential overpressurizatton events trust be considered when establishina the worst case event. ~3nme events may he prev"olvel hv protective interlncks or bv locking nut power, ibi ..r events should lie revieweil on an inclividual basis. If the int er lock / power loc.kout. is accept able , it can lie excluded from the analve.es provided the controls to prevent the event are

[,,' in the pl ml lechnical Specifications.

3. The y tem must meet the design requirements of IEEE 279 (see inolementation). The system may be manually enabled, huwever, the electrical instrumentation and control system must provide alarms to alert the operator to:

a. properiv enable the system at the correct plant condition luring cooldown,

b. indicate if a pressure transient is occurring.

4. To as,ure operational readiness, the overpressure protection system must be tested in the following manner:

a. i W.t mu .t be perfo'rmeet to av.oro operability of the sys tem alectronics prior to each shutdown.

b. A to.L fei valve operability must, as a minimum be ccnducted as specified in the A511E Code Section XI.

c. 'un equent tn system, valve, or electronics waintenance, a tr.t

.in that portion (s) of the system must be performed prior to declaring the system operational.

e

.I

4 ENCLOSURE 1

5. The e.y . ten mus t meet the req i' ments nf Regulatory Guide 1.26,

" Quality Group Classification, and Standards for Water , Steam ,

and na dioactive-Waste-Containing Components of Nuclear Power Plants" -

and Section lll of the ASME Code.

6. The overpressure protection system must be designed to function during an Operating Basis Earthquake. It must not comproinise th'e desion criteria of any other safety-grade system with which it would interface, such that the requirements of Regulatory Guide 1.29, " Seismic Design _ Classification" are met.

7. The overpressure protection system must not depend on the availability of of fsite power to perform its function.

8. Overpressure protection systems which take credit for an active component (s) to mitigate the consequences of an overpressucitation event must include additional analyses considering inadvertent system initiation / actuation or prov.ide justification to show that existing analyses bound such an event.

C. Impleyen ta tion The BrarC /echnical Position, as specified in Section B will be used in the review of all Preliminary Design Approval (PDA), Final Design

[',,' Approval (FDA), f4anuf acturing license (ML), Operating License (OL), and Construction Permit (CP) applications involving plant designs incorporatir.

pressurized water reactors. All aspects of the position will be appi u abi.

to al' arplications, including CP applications utilizing the replication ortion of the Connission's standardi:ation rogran, that are docketed a f ter !' arch 14, 1978. All aspects of the position, with the exceotion of enonable and justified deviatinns rrom IEEE 279 requirements, will be arplicable to CP, OL, ML, PDA, and FDA applications docketeti prior tn 'larth 14 1978 but for which tne licensina action has not been cnmplet nd as of liarch 14, 1978. Holders of appropriate PDA's will be info w'd by letter that all aspects of the position with the exception of IEEE 279 will be applicable to their approved standard designs and that such designs should be modified, as necessary, to conform to the position. Staf f approval of proposed modifications can be applied for either bv application by the FDA-holder on the PDA-decket or by each CP applicant referencing the standard design on its docket.

The tollowinei q u i <le l i ne r, may be used, if necese,ary, to alleviate impacts on lit. n ino e.che lules for plant? nvolved in licensing proceedings nearing completion on March 14, 1978:

I I.

m

. ENCLOSURE 1

1. Those applicants issued an r" . iring the period between March 14 1978 and a date 12 months thereafter may merely commit to meetino the pnsition prior to OL issuance but shall, by license ccndit:6n, i

be required to install all required staff-approved 43011/ . . .. 1.sn :

l 3

prior to plant startup following the first scheduled rs(ueling i outane.

e i I 2. Those applicants issued an OL beyond March 14, 1979 shall install all required staff-approved modifications prior to initial plant i'.a r tup .

3. Those applicants issued a CP, PDA, er ML during the period between fiarch lit, 1978 and a date 6 months thereaf ter may merely coimtit 3

to meeting the position but shall, by license condition, be reautred to amend the aoplication, within G months of the date of

, issuance of the CP, PDA, or ML, to include a description of the prnposed modifications and the bases for their design, and a request for staff approval, i

, 4 Those applicants issued a CP, PDA, or ML after September la, 1978 shall have staff approval of proposed modifications prior to issuance of the CP, PDA, or ML.

l D. Re ferences

1. NtlREr,-0138, Staff Discissien of Fifteen Technical Issues Listed in M.tachment to November 3,1976 Memorandum from Director, ilRR, i to flRR Staff.

l l .

1

, e D - ___ _ _ _ _ _ _ _ __ _ _ _ _ . - - - - - - - - - - - -_

ENCLOSURE 2

~ -

DOCUMENT NUMBER REVISION TITLE l

RG 1.27 2 Ultimate Heat Sink for Nuclear Power Plants

. RG 1.52 1 Design, Testing, and Maintenance Criteria for Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light Water Cooled Nuclear Power Plants

• RR 1.59 2 Design Basis Floods for Nuclear Power Plants

• RG 1.63 2 Electric Penetration Assemblies in Containment Struct es for Light Water Cooled Nuclear Power Plants

. RG 1.91 1 Evaluation of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plant Sites RG 1.102 1 Flood Protection for Nuclear Power Plants

. RG 1.105 1 Instrument Setpoints

, RG 1.108

• 1 Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants

• RG 1.115 1 Protection Against Low-Trajectory Turbine Missiles RG 1.117 1 Tornado Design Classification

• RG 1.124 1 Service Limits and Loading Combinations for Class 1 Linear Type Component Supports

. RG 1.130 0 Design Limits and Loading Combinations

~

for Class 1 Plant- and Shell-Type Component Supports l _ --_-_ _____ __________ _____ .

l l

l ENCLOSURE 2 (CO! TINUED)

DOCUMENT NUMBER REVISION TITLE RG 1.137 0 Fuel Oil Systems for Standby Diesel Generators (Paracraph C.2)

. RG 8.8 2 Information Relevant to Ensuring that occupational Radiation Exposures at Nuclear Pcwer Stations Will be as Low as is Reasonably Achievable (Nuclear Power Reactors)

BTP ASB Guidelines for Fire Protection for Nuclear 9.5-1 Power Plants Under Review and Construction RG 1.141 0 Containment Isolation Provisions for Fluid Systems

4 ENCLOSURE 3

1. Environmental Control and Qualification Outside Containment - qualification review item E.27

2. Containment Response to Steam Line Break Inside Containment - qualification review item E.28

3. The Analysis of Asymetric Loads on Components and Supports Located Within Containment Subcompartments - qualification review item E.29

4. Containment Leakage Test Program to Meet Appendix J to 10 CFR Part 50 -

qualification review item E.30

5. Single Failure Evaluation of Postulated Steam Line Break Accidents -

qualification review item E.31

6. Effects of Abnormal Grid Voltage - qualification review item E.32

7. Cooling Water Supply to Reactor Coolant Pumps - qualification review item E.33

8. Fuel Rod Bowing - qualification review item E.34

9. Ductility of Reinforced Concrete and Steel Structural Elements Subjected to Impactive or Impulse loads - SRP Section 3.5.3

10. Response Spectra in Vertical Direction - SRP Section 3.7.1

11. Air Blast Loads - SRP Section 3.8.4

12. Tornado Missile Impact - SRP Section 3.5.3

13. Passive Failures During Long-Term Cooling Following LOCA - SRP Section 6.3

14. Control Room Position Indication of Manual (Handwheel) Valves in ECCS - SRP Section 6.3

15. Long- Term Recovery from Steamline Break: Operator Action to Prevent Overpressurization - SRP Section 15.1.5

16. Pump Operability Requirements - SRP Sections 5.4.6, 5.4.7, and 6.3

17. Gravity Missiles, Vessel Seal Ring Missiles Inside Containment - SRP Section 3.5.1

18. Design Guidelines for Water Hammer in Steam Generators with Top Feedring -

SRP Section 10.4.7

- - - - _