ML19209A569
| ML19209A569 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 09/14/1979 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19209A570 | List: |
| References | |
| NUDOCS 7910050006 | |
| Download: ML19209A569 (19) | |
Text
DISTRIBUTION ggp y 4 )379 Docket File M. Collins, 016 NRC PDR D. Ross Local PDR R. Tedesco LWR #3 File S. Pawlicki NRR Reading R. Bosnak H. Denton F. Schauer E. Case K. Kniel D. Vassallo T. Novak D. Ross Z. Rosztoczy S. Varga V. Benaroyc J. Stolz W. Butler R. Baer Chief, ICc3
- 0. Parr V. Moore R. Rubenstein R. Vollmer F. Williams M. Ernst R. DeYoung F. Rosa R. Mattson W. Ganmill C. Miles, OPA P. Leech IE (3)
M. Duncan J. Souder W. Regan MPA G. Chipman OELD J. Collins D. Skovholt W. Kreger A. Dromerick G. Lear M. Rushbrook B. Youngblood E. Reeves J. Stepp C. Parrish L. Hulman A. Schwencer R. Diggs D. Eisenhut D. Sells C. Stephens D. Muller E. Pleasant J. Knight H. Berkow R. Denise D. Lanham, 016 F. Schroeder M. Jinks (3)
(53 gg 1108 218 7f10050OOb f
- pa arcuq'o UNITED STATES I f ','c. <f(',h NUCLEAR REGULATORY COMMISSION 7 J, '
W)/, E WASHINGTON, D. C. 20555 k.... f SEP 141979 Docket No.
50-338 MEMCRAriDUM FOR:
D. Eisenhut, Acting Cirector, Division of Operating Reactors, NRR FROM:
D. Vassallo, Acting Director, Divisic of Project Management, NRR
SUBJECT:
TRANSFER OF THE NORTH ANNA POWER STi. TION, UNIT 1 TO OPERATING REACTORS BRANCH NO. 1 Effective on the date of this memorandun the project management responsibility for the North Anna Power Station, Unit 1 is transferred from LWR Branch No. 3, Division of Project Management to Operating Reactors Branch No.1, Division of Operating Reactors.
The licensee, the Virginia Electric and Power Company, received a license (NPF-4) on November 26, 1977, which authorized Unit.1 to load fuel and maintain the reactor in a cold shutdown condition.
A chronology for the issuance of the Amendn. ants to license NPF-4 and the date of authorization is presented below. Also a brief description is provided of the Amendments presently issued to License NPF-4.
C hronolocy Amendment No.
Date of Issue Description 1
January 25, 1978 Authorized Unit 1 to be operated in a hot standby condition 2
March 17, 1978 Exempted Unit 1 from certain technical specifications relating to the recirculation spray pumps while in the hot standby condition 3
April 1,.978 Authorized full power operation at 2775 Mwt 1108 219
D. Eisenhut SEP 161979 Ar:endrent No.
Date of Issue Description 4
May 8,1978 Deleted two conditions from Amendment 16. 3 to NPF-4.
These conditions were
- 2. D. (3)a and 2. D. (3)n and related to tests and procedures to be performed by the licensee prior to achie'. ng f ull power.
5 May 19,1978 Appr ved the licensee's proposed design modifications for the final solution to the NPSH problem for the recirculation spray punps.
The amendment also changes Appendix A to the Technical Specifications related to the heat flux hot channel factor (Fq, limit).
6 June 23,1978 Revised Technical Specification 4. 5. 2.f.2 to specify a minimum acceptance pressure at recirculation flow for the low-head safety in.jection pumps of greater tnan or equal to 56 pounds per square inch gauge 7
July 3,1978 Revises Condition 2.D(3)j to permit the licensee to submit environmental qualification testing results of Barton transnitters prior to October 1, 1978.
8 March 6,1979 Approved North Anna Power Station Units 1 and 2 Fire Protection Program.
9 February 23, 1979 Approved Security Plan for North Anna in accordance with 10 CFR 73. 5 5.
1i08 220
D. Eisenhut SEP 141979 Amendment No.
Date of Issue Description 10 April 27,1979 Required Flow Splitter Plate Surveillance.
11 June 14,1979 Approved VEPC0's re-organization and
- attery Surveillance Program.
12 June 28,1979 Authorized a change in allowable settlement limits - Class 1 structures.
13 August 3,1979 A one-time extension of certain surveillance fre-quencies in Appendix A Technical Specifications to September 15, 1979.
14 August 17, 1979 Auth'orized an increase in fuel storage capacity.
The current status of items requiring further staff actions and the organization responsible for completing these items are identified in Enclosure 1.
Lists of generic problems and Regulatory Guides used during the licensing review, with reference to the locations where relevant information or evaluations of record may be found, are included in Enclosures 2 and 3, respectively. is a DSE memorandun dated July 18, 1979 which sunmarizes the environmental status of the project and transfers the environmental project management responsiblity from DSE to DOR. Enclosure 5 is the service list for this plant.
By copy of this meno, DSE, IE, MPA, OELD, Records Facility Branch, ADM, Public Affairs and Docketing and Service Branch, Secy, are being notified of the following safety personnel changes which are to be effective on the date of this memorandum.
i108 221
D. Eisenhut SEP 141979 FROM 0
Project Manager A. W. Dromerick A. Schwencer Branch Chief
- 0. D. Parr A. Schwencer Acting Assistant Director S. Varga W. Garmill Licensing Assistant M. Rushbrook P. Kreutzer Environ, ental personnel changes described in Enclosure 4 are as follows:
FROM T0_
Environmental Project Manager P. Leech A. Schwencer Acting Branch Chief D. Sells A. Schwencer Acting Assistant Director W. Regan W. Gannill Licensing Assistant M. Duncan P. Kreutzer D. B. Vassallo, Acting Director Division of Project' Management Office of Nuclear Reactor Regulation
Enclosures:
As Stated 4
1108 222
ENCLOSURE 1 CURRENT STATUS OF ITEPS REQUIRING STAFF ACTIONS NORTH ANNA POLER STATION UNIT 1 00CKET NO. 50-338 FACILITY OPERATING LICENSE NPF-4 The items requiring further staff action are as follows:
1.
Overpressurization P rotection for the R eactor Coolant System Itera 2.D (3)b of Amendment 3 to NPF-4 requires the '.censee to install a long term means of protection against reactor ce lant system
'verpressurization prior to startup following the first refueling
.;h ut d o m. The licen ee is a member of a utility 3roup that is developing a lont,-tern solution to mitigate the consequences of pressure transients during water - solid operation. The design modification being considered utilizes the power - operated relief valves to preclude violating Appendix G limits (see licensee's letters of April 14, 1977 and June 19, 1977 and Section 5.2.8 of Supplement No. 7 to the SER).
Evaluation will be by the Civision of System Safety (cognizant DSS reviewers S. Israel and Glen Kelly).
In a letter dated December 21,1978, the licensee provided the information specified in request 5.79 in our letter dated January 12, 1977.
In our letter of Parch 21, 1979 we advised the licensee that we require additional information. VE PC0' s ubmitted additional information on April 17 and 23,1979.
Panagement respon-sibilities will be carried out by Operating Reactor Branch No.1.
2.
Reevaluation of Fire P rotection P rocram Item 2.D(3)f of Amendment 3 to fSF-4 requires the licensee to implement prior to startup following the first regularly scheduled refueling outage, the administrative procedures, controls and the fire brigade program associated with the fire protection program.
With respect to our letter dated September 30, 1976, the licensee responded with submittals on April 1,197 7, September 7,1977, December 22, 1977, and January 3, 1978.
Evaluation is being conducted by the Division of Systems Safety (cognizant DSS reviewers - P. Patthews and G. Harrison).
The staf f's letter of May 12, 1978 transmitted questions and staff positions to the licensee based on the above licensee submittals.
A meeting concerning the matter was held with VEPC0 on July 25 and 26,1978.
The licensee formally responded to our Fay 12, 1978 request on September 29, 1978.
DSS completed thei; review on December 15, 1978.
Amendment No. 8 regarding this matter was issued Parch 6,1979.
P00RBnyAt 1108 223
. 3.
Item 2.D(3)g of Amenument 3 to f1PF 4 required that the licensee submit for our review within six months of issuance of Amendment 3 (1) the technical specification trip setpoint values and (2) the technical specification allowable values.
In a letter dated Parch 15, 1977 we requested that VEPC0 provide information regarding trip setpoint v al ues.
In a letter dated Pay 25,1977 VEPC0 advised that they would provide trip setpoint values six months after issuance of an operating license. This information was submitted on September 29, 1978.
Evaluation and management of this matter is to be performed by the Division of Operating Reactors.
4.
Item 2.D (3)h of Amendment 3 to f PF-4 requires VEPC' to install qualified stem mounted switches for 21 in contain:...nt isolation valves prior to startup following the first regularly scoeduled refueling o ut age.
Implementation of this matter will be verified by the Cf fice of Inspection & Enforcement.
Vanagement responsibility will be carried out by Operating Reactors Branch tb.1 (see Section 3.10.3 cf Supplement tb. 9 to the SER).
5.
Item 2.D(3 )i. of Amendment 3 to fPF-4 requires that prior to the startup following the first regularly scheduled refueling outage, VEPC0 shall install and have operational the area ambient temperature monitoring system outside containment.
Amendment fb. 5 to NPF-4 revised this condition in order to correct implementation aate. Implementation of this matter will be verified by the Office of Inspection and Enforcement. Management responsibility of this system will be carried out by the Division of Operating Reactors.
6.
In item 2.0(3)j of Amendment 7 to fPF-4 we require that VEPC0 provide prior to October 1,1978 the result of proper sequential qualification testing performed for (1) Barton 386/752 (now designated Barton 764 and Barton 393) (now designated Barton 763) transmitters.
In a letter dated August 22, 1978, and June 22, 1979, the licensee provided additional information regarding this matter.
Evaluation will be performed by the Division of System Safety and management responsibility will be carried out by the Division of Operating Reactors (TAC 5025, cognizant DSS reviemrs F. Orr and D. Thatcher). Based on our review of August 22, 1978 letter, RSB indicates instruments were acceptable for use on tbrth Anna for one full power year.
Instrumentation and Control has given a conditional approval for use in fiorth Anna. RSB has not completed its review of the information provided in the licensee's letter of June 22, 1979.
Instrumentation and Control expects to complete its final review by October 15, 1979.
il08 224
. 7.
In item 2.0 (3)k of Amendment 3 to NPF-4 we required confirmatory tests of the outside recirculation spray pumps. In a letter dated June 2,1978 VEPC0 submitted the results of this testing, and we are reviewing this inf ormation. The Division of System Safety will complete the evaluation and the Division of Operating Reactors will carry out the management responsibility.
(TAC 4904, cognizant DSS revietee - J. Olshinsky and S. Israel).
The estimated completion date for our review is September 30, 1979).
8.
In item 2.D(3)k of Amendment 3 to PPF-4 we required long-term confirmatory testing of the low-head safety inject 4 ;r, punps. The licensee has completed this testing and submitted report concerning this r:atter on July 14, 1978. The Division of Sys'.en Safety is reviewing the test information which was submitte.1 by the licensee and the Division of Operating Reactors will carry out the management responsibility. (TAC 4904, cognizant DSS revievers - J. Olshinsky and S. Israel). The estimated completion date for our review is September 30, 1979).
9.
In item 2.D(3 )m of Amendment 3 to NPF-4 we required the licensee to submit by P.ay 1,1978, a vibration modal analysis of the inside recirculation spray pumps. In April 1978, the licensee submitted a modal analysis.
Division of System Safety is pr.esently reviewing this information and will provide Safety Evaluation to the Division of Operating Reactors.
The Division of Operating Reactors will carry out management responsibility. (TAC 4904, cognizant DSS reviekers -
J. Olshinsky and S. Israel).
The estimated completion date for our review is September 30, 1979.
10.
In a letter dated Fay 1,1978 VEPC0 requested a technical specification change concerning the expansion of the fuel pool racks.
The change has been prenoticed and there are two petitions to intervene and a safety hearing will be held.
Review responsibility will be carried out by the Division of Systems Safety and the Division of Site Safety.
Vanagement responsibility will be carried out by the Division of Project Management. The Safety Evaluation was issued on January 29, 1979.
In the Safety Evaluation Report, heavy loads near spent fuel were discussed and found acceptable. The hearing will be held on August 14, 1979.
(TAC 4935, cognizant DSS and DSE reviewers J. Wermiel, S. Chan, C. Ferrell, T. Murphy and P. Leech).
11.
In a letter dated June 13,1978, VEPC0 requested that the technical specification concerning sers ice water pump house settlement limit be changed from 0.15 feet to 0.33 feet.
Review responsibility was carried out by the Division of System Safety and the Civision of Site 1108 225
. Safety and Environmental Analysis.
Panagement responsibility was carried out by the Division of Project Management. Amendment No.12 regarding this matter was issued June 28, 1979 and this matter is considered complete. Hov.ever, if the Appeals Board issues an order that requires additional work on this item, the Division of Systems Safety will have the review responsibility and the Division of Project Fanagement will have the management responsibility.
12.
In the ACRS letter of January 17, 1977, the ACRS recommended that the fiRC staff review the seisnic design to assure itself that significant seismic design margins exist in all systems required to accomplish safe shutdov:n of the reactors and continued shutdown heat removal, given an SSE.
In a letter dated June 14, 1978 we advised t!
ACRS that we would issue a report regarding this matter in Auge,t 1978.
The seismic evaluation of the systems and preparation of the eport will be performed by the Division of Systems Sa#ety.
The Division of Operating Reactors will carry out management responsibility (cognizant reviewer K. Desai -
estimated completion date November 30, 1979).
13.
By NRC letter dated October 17, 1977 concerning the inservice inspection and testing program the licensee was granted written relief from the requirements of Section XI of the ASME code for pumps and valves testing program in accordance with Technical Specifications 4.0.5a.
On December 16,1977, VEPC0 submitted a letter o'utlinin.g the'r initial inservice inspection program for the reactor coolant system.
They further amended their request by letter dated October 27, 1978. Corm:ercial operation for Unit 1 was initiated in May 1978. DSS will evaluate the licensee's submittal. ( Cognizant DSS review M. Hum). Ma nagement responsibility will be carried out by the Division of Operating Reactors.
The estimated completion date for our review is fbvember 30, 1979.
14.
In a letter dated July 7,1978, VEPC0 responded to our generic letter of May 13, 1978 mgarding diesel generator alarms at the fiorth Anna Power Station Units 1 and 2.
The evaluation has been completed and a letter advising the applicant that his design is acceptable was transmitted January 18, 1979.
15.
In a letter dated August 14,1978, VEPC0 submitted in accordance with our request of June 14, 1978 changes to Technical Specifications
- 4. 8.1.1. 3 a nd 4. 8.1. 2.
The proposed changes identify additional surveillance requirements for the diesel generator batteries.
Evaluation was carried out by the Division of Systems Safety.
(TAC 5028, cognizant DSS review D. Thatcher).
/cendment tb.11 related to this raatter was issued June 22, 1979.
16.
In letters dated August 31, 1978, Decembe r 26, 1978 and March 8,1979 VEPCO requested changes in Technical Specification ( Appendix A) 6.4.1 and Environmental Specification 5.1 related to VEPCO's reorganization. Amendment No.11 related to this matter was issued June 22,197 9.
i uos 226 P00R OR E l
. 17.
In a letter dated November 21,1978 VEPC0 requested an Amendment to Technical Specifications 3. 7.1.6 and 4. 7.1.6 related to cation conductivity limits and surveillance requirements for the secondary system. A technical assistance request was issued to DSS on November 21, 1978.
(TAC-50827).
Evaluation is being performed by B. Turovlin. DDR will have management responsibility.
16.
In a letter dated November 2,1978, VEPC0 requested relief from certain requirements of ASP.E XI for Inservice Testing of Certain pumps and valves. A technical assistance request was issued November 29, 1978.
(T AC -5085 ). Evaluation will Le performed by K. De:ii and DOR will have management responsibility.
19.
In letters dated tarch 6,1979, and July 3,1979, VEPC0 submitted Technical Specification to implement Appendix I to 10 CFR 50 in accordance with Division of Project Management letter of July 10, 1978.
Eval uation will be performed by J. Collins /J. Eoegli. D0R will have management responsibility (TAC-8147 ).
20.
In a letter. dated Parch 23, 1979, the licensee advised that VEPC0 amended the security program.
Division of Operating Reactors (reviewer -
C. Gaskin) will perform the review of this amendment and Operating Reactors Branch will have management responsibility.
A letter regarding the contingency plan was transmitted to the licensee on July 10, 1979.
21.
In a letter dated July 6,1979, VEPC0 requested a change in the technical specifications related to the single dropped rod event. A technical request is being prepared. DSS will perform the evaluation and the Division of Operating Reactors will have management responsibility. No schedule has been established.
1108 227
ENCLOSURE 2 CURRENT STATUS OF GENERIC REVIEW ITEf5 NORTH ANNA P0kER STATION, UNIT 1 A.
Items which have been resolved and require no further NRC action.
Item R eference 1.
Containment leak testing Appendix J SER p 6-8 and p 6-9 TS p 3/4 6-2 through 3/46-5 2.
FAC ECCS Evaluation Sr < p 6-11 through 6-13 S ER No.1 p 6-3 and p 6-4 SSER t,'o. 2 p 6-1 a nd p 6-2 SSER No. 3 p 6-2 and p 6-3 SSER No 4 p 6-1 and p 6-2 SSER No. 8 p 6-10 and p 6-11 SSER fio. 9 p 6-11 through 6-13 Amendment No. 5 to NPF-4 Item A of Safety Evaluation 3.
Emergency Planning SER p 13-2 through 13-5 SSER No. 2 p 13-1 4.
Filter Technical Specifications TS p 3/4 7-21 through 3 /4 7-25 5.
Flooding of safety equipment SER p 10-3 SEP, p 3-4 S SER No. p 2-1, p 2-2, p 3-1 and p 3-2 6.
High energy line break SER p 3-5 through 3-7 7.
QA program for operation SER p 17-1 through 17-5 P00RBRGNAL 1108 228 Item Reference 8.
Asymetric LOCA loads SER p. 3-13 through 3-17, p. 4-9 and p. 4-10 SSER No. 7 p. 3-1, 3-2, 4-1, 4-2 and 4-3 9.
Snubbers TS p. 3/4 7-28 through 3/4 7-67 10.
Steam generator feedwater - flow SER p.10-2 and 3 instability Amendment No. 4 to NPF It m A of Safety Evaluation
- 11. Resctor vessel surveill w.e S~R p. 5-7 and p. 5-8 program Appendix H 3 p. 3/4 4-26 and p. 3/4-24 12.
Secondary Water Chemistry SER p. 5-6 Monitoring requirements TS 3/4 7-11 and 7-12 (see item 17, Enclosure 2) 13.
ECCS reevaluation to account SSER No. 3 p. 6-2 and p. 6-3 for increase vessel head fluid temperature 14.
Fuel rod bow effects on hot SER p. 4-15 and p. 4-16 channel enthalpy rise factor SSER No. 3 p. 4-1 TS p. 3/4 2-9 15.
Boron dilution accident Licensee responded to this concern in Amendment 44 of FSAR - p. 6.4-1 through
- p. 6.4-3 SER p.15-4
- 16. Qualification of radiation TS p. 6-5 protection manager 17.
Fuel handling accident inside SSER No. 7 p.15-1 through containment
- p. 15-4 18.
HPSI - LPSI flow stops TS p. 3/4 5-4 and p. 3/4 5-5 F
E00R ORGIML 1108 229 I t er.
R eference 19.
Fuel cast drop protection SER p. 9-3 and p. 9-4 system SSER p. 9-1 and p. 9-2 20.
Primary component supports SSER fb. 3 p. 5-1 through (fracture toughness)
- p. 5-8 SSER No. 6 p. 5-1 and
- p. 5-2 21.
Reactor vessel overpressurization:
E ;losure 1 - Item 1 of short term program only t '.is letter B.
Itens which have been evaluated in light of current NRC requirements /
guidance and for which measure that will make the status acceptable have been initiated by the licensee.
Item R eference none none C.
Ite..is which have been evaluated in light of current fRC requirement /
guidance and which are presently unresolved.
Further action by NRC and/or the licensee may be required in the future. '
I tem R eference 1.
Seismic qualifications of class See Item 6 of Enclosure 1.
IE Electrical ~ and Pechanical Equipment 2.
Implementation of 10 CFR 50-55a (9,i See Item 13 of Enclosure 1.
inservice inspection program 3.
Implementation of 10 CFR 73. 55 Plan approved for issuance of operating license ( see fPF-4 Amendment 3 Section 2E ).
Final implementation now being reviewed by staff. Request for additional information transmitted to licensee on July 21, 1978.
Ame n d-ment 9 to NPF-4 regarding this matter was issued.
4.
Reactor vessel overpressur ization See Item 1 of Enclosure 1.
protection long-term requirements J
j
]
. Item R eference 5.
Fire protection See Item 2 of Enclosure 1.
6.
Diesel generator lockout NRC letter to licensee transmitted May 31, 1978.
Licensee's response submitted July 7,1978.
Evaluation completed (see Item 14 ).
7.
Pressure vessel fracture DP' letter concerning toughness properties piassure vessel fracture tcughness properties transmitted to licensee Decembe r 6, 197 7.
Licensee provided information in a letter dated December 11, 1978.
DSS considers this issue resolved and that Sections 5.2 & 5.3 cf the SER still applies.
8.
Degraded grid voltage License'e submitted report on December 10, 1976.
Position transmitted to licensee on July 28, 1978.
Evaluation complete - lettar to licensee March 20, 7 9.
Appendix I ( ALARA)
SSER Np. 2 p.11-1 through p.11-13 letter transmitting new technical specifications sent to licensee on July 10, 1978.
Licensees response submitted Parch 6,1979 and July 3,197 9.
(See Enclosure 1, Item 20).
D.
Items which have been evaluated in light of current NRC requirements /
g idance and which are unacceptable.
Itec R eference none none P00R OREL 1108 231
- E.
Items which have not been evaluated in light of current NRC requirements /
g uida nce.
Item R eference none none F.
Items which are not applicable to the fbrth Anna Powr Station, Unit 1.
All items identified as specific to GE, CE, or B a d W plants.
9 9
1108 232
ENCLOSURE 3 R ecLlatory G uides U sed During L icensing R eview Guide Nurber Title R eference Chapter 2 1.23 Cnsite Meteorological Programs ( February 1972)
SER - Section 2.3 SSER No.1 Section
?. 3. 3 1.4 Assunctions used for SER - Section 2.3 evaluating the potential radiological consequences of a loss-cf-coolant accident for pressurized water reactors (Revision 2, June 1974).
1.27 Ultimate heat sink for nuclear SER - Section 2.4 power plants.
1.59 Design Basis floods for nuclear SSER No. 2 pl ants.
Section 2.4 Chapter 3 1.46 Protection against pipe whip SER - Section 3.6 inside containment.
1.61 Damping values for seismic SER - Section 3. 7 design of nuclear power plants.
1. 60 Design response spectra for SER - Section 3. 7 seismic design of nuclear power plants.
1.12 Instrumentation for earthquakes SER - Section 3. 7 1.18 Structural acceptance test for SER - Section 3.8 concrete primary reactor contain tment.
1.48 Design limits and loading SER - Section 3.9 combinations for sei saic Category 1 fluid systems components.
1.67 Installation of overpressure SER - Section 3.9 protection services.
1i08 233 Guide Humber Title R eference Chapter 4 1. 20 Comprehensive vibration SER - Section 4.2 assessment program for reactor internal during preoperational and initial startup testing.
1. 48 Design limits and loading SER - Section 5.2 combinations for seismic Category 1 fl uid systems components.
1.67 Installation of overpressure SER - Section 5.4 protection devices Chapter 5 1.14 Design limits and loading SER - Section 5.4 combinations for sei saic Category I fl uid systems components.
Reactor Coolant Pump flywheel integrity.
C hapte, 6, 1.7 Control of ccmbustible gas SER - Section 6.1 concentratior, in containment following a loss-of-coolant accident.
1.11 Instrument lines penetrating SER - Section 6.2 primary reactor containments.
1.79 Preoperation testing of ECCS SSER No. 5 Section 6. 3.4 SSER No. 8 Section 6. 3.4 Chapter 7 1.6 Independence between redundant SER - Section 7.0 standby (onsite) power sources and between their distribution syst ems.
P 1.7 Control of combustible gas SER - Section 7.0 concentrations in containment following a loss-of-coolant accident.
1108 234 Guide Nucter Title R eference 1.9 Selection of diesel generator SER - Section 7.0 set capacity for standby power suppl i es.
1.11 Instrument lines penetrating SER - Section 7.0 primary reactor containment.
1.22 Periodic testing of protection SER - Section i.
system actuation functions.
1. 32 Use of IEEE Std 308-1971 critr ia SER - Section 7.0 for class IE electric system:
for nuclear poter generating systems.
1.40 Qualification tests of continuous SER - Section 7.0 duty motors installed inside the containment of water-cooled nuclear power plants.
1.41 Preoperational testing of redundant 5LR - Section 7.0 on-site electric power systems to verify proper load group assignment:,.
1.47 Bypassed and inoperable status SER - Section 7.0 indication for nuclear power plant safety systems.
1. 53 Application of the single SER - Section 7.0 failure criterion to nuclear pov.er plant protection systems.
1.62 Panual initiation of protective SER - Section 7.0 actions.
1.63 Electric penetration assemblies SER - Section
- 7. 0 in containment structure for water-cooled nuclear pomr plants.
Chapter 8 none il08 US
_4 G uide Nur.ber Title R eference Chapter 9 1. 29 Seismic design classification.
SER - Section 9.1 1.13 Fuel storage facility design SER - Section 9.1 basis.
1.27 Ultimate heat sink for nuclear SER - Section 9. 2 power plants.
Chapter 10 1. 29 Seismic Design classifications.
SER - Section 10. 2 SSER No. 2 Section 10.2 Chapter 11 1.109 Calculation of annual average doses SSER No. 2 to r.ian from rcutine release of Section 11.1 reactor effl uents for the purpose of implementing Appendix 1.
1.21 Measuring, eval uating and reporting SER - Section 11.5 radioactivity in solid-wastes and releases of radioactive materials in liquid anu gaseous effluents from light-water-coolant nuclear power Chapter 12 none Chapter 13 1.8 Personnel selection and training SER - Section 13.1 1.10 1 Emergency Planning SER - Section 13.3 1.33 Quality Assurance program SER - Section 13.4 requirements (operation) 1.17 Protection of nuclear power SER - Section 13. 6 plants against industrial sabotage i108 236
. Guice Number Title R eference Chapter 14 1. 68 Preoperational and initial starting SER - Section 14.0 test programs for water' cooled poker reactors.
Chapter 15.0 1.4 Rev.1 Assumptions used for calculating SER - Section 15.4 the radiological consequences of a loss-cf-coolant accident for pressurized water reactors.
1.77 Assumptions used for eval uating SER - Section 15.4 a control rod ejection accident for pressurized water reactors.
i108 237