ML18022A444

Forwards marked-up Amend 37 to Fsar.Justification for Revs to Listed Chapters Also Encl.Revs Do Not Alter Conclusions in SER Through Sser 3 (NUREG-1038).Amend Will Be Formally Reissued in Public Form

ML18022A444
Person / Time
Site:	Harris
Issue date:	10/03/1986
From:	Cutter A CAROLINA POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-1038 NLS-86-381, NUDOCS 8610140189
Download: ML18022A444 (164)

Text

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REQULATORY INFORNATION DISTRIBUTION 'SYSTEH ACCESSlON NBR: $ 610140189 DOC. DATE: $ 6/10/03 NOTARIZED: YES DOCKET ¹ FACIL: 50-400 Sheav'on Harv is Nuclear Power Planti Unit 1~ Cav olina 05000400 AUTH. NANE AUTHOR AFFILIATION CUTTER> A. B. Cav o1 ina Powev Zc Light Co.

RECIP. NANE RECIPIENT AFFILIATION DENTONi H. R. Office of Nucleav Reactov Regulation~ Div ectov'post 851125 SUB JECT: Fov'wav'ds mav ked-up Amend 37 to FSAR. Justification f ov'evs listed chaptev's also encl. Revs do not altev'onclusions in SER thv ough SSER3 (NUREQ-103$ ). Amend will be formal lg reissued in public form.

DISTRIBUTION CODE: B001D COPIES RECEIVED: LTR ENCL SI ZE:

TITLE: Lic ens ing Submi tta 1: PSAR/FSAR Amdt 5 5 Related Cov'res p ond en ce NOTES:Application fov pevmit v enewal filed. 05000400 RECIPIENT COPIES RECIPIENT COP IES ID CODE/NANE LTTR ENCL ID CODE/NANE LTTR ENCL PWR-A EB 1 1 PWR-A EICSB 2 2 PWR-A FOB 1 1 PWR-A PD2 L* 1 PWR-A PD2 PD 1 BUCKLEY'S B 01 2 2 PWR-A PSB PWR-A RSB 1 INTERNAL: ADN/LFllB 1 0 ELD/HDSi 1 0 IE FILE 1 1 IE/DEPER/EPB 36 1 1 IE/DGAVT/GAB 21 1 NRR BWR ADTS

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@MD, Carolina Power & LIght Company SERIAL: NLS-86-381 Mr. Harold R. Denton, Director OCT 3 Lgge Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NO. l - DOCKET NO. 50-tI00 FSAR AMENDMENT37

Dear Mr. Denton:

Carolina Power inc Light Company (CPdcL) hereby submits a hand-marked copy of Amendment 37 to the Shearon Harris Nuclear Power Plant (SHNPP) Final Safety Analysis Report (FSAR). This amendment includes revision to Chapters 2, 3, 5, 6, 7, 9, ll, l2, 13, IO, and l5. The attached table provides justification for these revisions which constitute the known remaining changes which need to be made prior to issuance of an operating license. CPttcL has reviewed the changes in this amendment against the SHNPP Safety Evaluation Report (SER) (NUREG-1038) through Supplement 3, and it is our position that these changes do not alter the conclusions of the SER.

Each page bears the amendment number, and changes are indicated by vertical bars in the margin. This amendment is hand marked on current FSAR pages due to the press of time prior to licensing. It will be formally reissued in a published form shortly; therefore, an effective page list and instructions for entering the revised pages are not included with this submittal.

If you have any questions, please contact me.

Yours very tr y, A. B. Cutter, - Vice President Nuclear Engineering R Licensing ABC/3DK/bmc (50213DK)

Attachment cc: Mr. B. C. Buckley (NRC) 0e(

Mr. G. F. Maxwell (NRC-SHNPP)

Dr. 3. Nelson Grace (NRC-RII) S ill A; B. Cutter, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are officers, employees, contractors, and agents of Carolina Power dc Light Company.

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Also Available On CONTAINMEHT ISOLATION SYSTEM DATA ER+mj Aperture Card SHNPP FSAR CA@a PDlETRATION DATA VALVE DATA

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pp va4 ~aa SUPPLY IIIC 301 56 A NO YES INSTRUMENT V33 CK SA C C C NO YES YES .

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73 I S7 HYDROGEN 19 ANALYZER ipgggO V349 A GL SO A <60 P C C YES YES YES IRC aRC.

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105 56 HYDROGEN V309 GL SO A <60 C C YES YES 19 ANALYZER SP-VBN SP-VOBIS V315 GL SO A <60 C C YES YES YES 6.2.4 31b Amendment No~

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(",A.R9 N~PURGE LO SPAN CALIBRATION ir I1 REMOTE CONTROL PANEL A Also AvaBable On CLASS 1E HI SPAN GASES II Aperture Card II REMOTE SAMPLE M Hp hNhLYZER: DILUTION PANEL D RY N~ PURGE 73 S

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REACTOR AUXILIAR ILDINg CONTAINMENT 37.

AMENDMENT NO SHEARON HARRIS NUCLEAR POWER PLANT Carolina Power 5 Light Company FINAL SAFETY ANALYSIS REPORT POST ACCIDENT HYDROGEN MONITORING SYSTEM FIGURE 6.2.5-7

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SIINPP FSAR b) The sample valves, containment penetration and piping up to and including the outermost containment isolation valves for the sample feed header and sample return line are ASHE Section III, Safety Class 2, Seismic Category I and are designed to retain their integrity and operability under all conditions following a design basis accidents c) All materials and equipment required by this system are selected to be compatible with the environmental conditions anticipated during accident operation and are suitable for a lifetime consistent with that of the plant, d) The system samples containment air, providing the means to measure the containment air hydrogen concentration and to alert the operator in the event that a high hydrogen concentration is'etected, in accordance with the requirements of Regulatory Guide 1.7.

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e) All containment isoLation valves are normaLLy closed and faiL closed on loss of electrical power. Heans are provided to reopen valves, when required, after power is restored. In the event of a containment isolation signal, valves 2SP-V301 SA-1 and 2SP-V349SA-1 close and isolate containment penetration 73B. Valves 2SP-V300 SA-1 and 2SP-V348 SA-1 close to isolate penetration 73A. On power failure, all valves fail closed, insuring

~ 'isolation.

Co 'nt operation, Penetrations 73A and 738 serve a dual Eunction. During n enetrations permit, continuous sampling of xnment atmosphere for detect@ articulate iodine and gas airborne radioactivity. During acciden tions e penetrations permit the withdrawal of containment atmosphe r en sampLing. The radiation monitor is unqualified Eor -accident operation. is therefore isolated from the post"acc'ample stream by valves 2SP-V302 SB- , -V304 SA"1, and 2SP-V305 SA-1. With 'this arrangement, isoLation o 2SP-V303 OCA un uglified radiation monitors LLT-3502 is assured. The hydrogen anaL zer cabinet, tag number 1SP-7438-SA, is qualified for post-accident operation. The sampl.e Lane coming Erom and going to penetrations 73A and 738 respectively, contain only safety train A associated valves. Likewise, containment penetrations 86A and 868 use only safety train 8 associated valves on the hydrogen'analyzer sample Lines.

As a result, if one safety train Eails then the required redundancy for post-accident hydrogen sampling is still provided'f the associated valves fail to close when they should close, safety is not compromised since the hydrogen analyzer is qualiEied Eor post-accident operation, E) The Hydrogen Analyzer System consists of two identical units which pre completely independent oE each other and are powered from independent onsite sources. Therefore, assuming a single failure, process capability is available to monitor the hydrogen concentration in the Containment. See Table 6.2.5-7 which provides a failure modes and effects analysis.

g) The system is designed for remote-manual sampling capability with an intermittent cycle oE Hydrogen indication for six (6) diEEerent sample 6.2,5-3 Amendment No. + 37

SHNPP FSAR 6 ' ' CONTAINMENT LEAKAGE TESTING The Containment and containment penetrations are designed to permit periodic leakage rate testing in accordance with General Design Criteria (GDC) 52 and 53 and Appendix J to 10CFR50.

Testing requirements for piping penetration isolation barriers and valves have been established by using the intent of GDC 54, as interpreted in Appendix 'J to 10CFR50. Exceptions taken to Appendix J for Type A, B, or C tests are described and justified in Subsections 6.2.6.1, 6.2.6.2, and 6.2.6.3, respectively.

6.2.6.1 Containment Inte rated 'Leaka e Rate Test (T e A Test)

The design leakage rate for the Containment is 0.1 weight percent per day.

The actual leakage rate is tested and verified using the methods and requirements of Appendix J to 10CFR50 for Type A tests.

In accordance with Appendix J, a margin for possible deterioration of the Containment integrity during the service intervals between integrated leakage rate test (ILRT) is provided. The measured leak rate (L at peak test pressure) shall not exceed 0.75 of the maximum allowable value.

The structural integrity test (SIT) is conducted during the same test program as the preoperational peak pressure integrated leakage rate test. The SIT is conducted in conformance with the descriptions contained in Section 3.8.1 and with the exceptions taken to Regulatory Guide 1.18 as specified in Section 1.8. After the SIT peak pressur~~n, requirements and the. 37 containment stabilization at required pressurization are completed, the initial peak calculated pressure (P = 41.0 psig) e=e5=ea.

This sequence of testing is chosen to satisfy paragraph'I.F of Appendix J to 10CFR50, which specifies that the initial ILRT shall be conducted after, the Containment is completed and is ready for operation.

Subsequent peak calculated pressure tests are conducted as specified in Section 6.2.6.4.

Reduced pressure ILRT's (as described in paragraphsZII.A.4 andjjI.A.5 of 37 Appendix J to 10CFR50) are not performed during pre-operational testing or during periodic ILRT's. Industry experience has shown that extrapolation factors used to correlate the reduced and full pressure tests are not reliable and may be erroneous in some cases.

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~<s re ~ 6.2.6.1.1 Pretest Requirements g 2.( -Za A p uisite to the Containment integrated leakage rate test is th satisfactory etion of a series of local leakage tests s involves subjecting potential aths through the co 'ent boundary (i.e.,

containment penetrations) to sam' xtions occurring during the integrated leakage rate test. uctlng leakage tests allows discovery and elimination of 1 at s through the Containme 'out pressurizing the entire xnment structure. These local leakage tests e Type B ests described in Se ti n 6 2 6 2 nd 6.2.6.3.

6.2.6-1 37 Amendment No.

SHNPP FSAR The primary prerequisite for conducting an ILRT is a general inspection of the accessible interior and exterior surfaces of the containment structures and components to uncover any evidence of structural deterioration which may affect either the structural integrity or leaktightness of the Containment.

If there is evidence of structural deterioration, Type A tests shall not be performed until corrective action is taken in accordance with repair .

procedures, nondestructive examinations, and tests as specified in the applicable code specified in 10CFR50.55a.

6.2.6.1.2 Valve Positioning for the ILRT The containment isolation valves are positioned accompanying adjustments. Normal, LOCA, and ILRT positions for each isolation valve are shown on Table 6.2.4-1'.2.6.1.3 System Preparation for Type A Tests Systems are properly isolated, drained, or vented to reflect their worst potential status'following a LOCA to assure that the Type A test results accurately reflect the most restricting LOCA conditions. Systems required to maintain the Unit in a cold shutdown condition are operable in their normal mode and are not vented or drained. However, any of these system penetrations that require Type Clocal leakage tests as defined in Section 6.2.4 have the results of the local leakage tests added to the result. of the Type A test.

Systems used during the Type A test for sensing the leakage are not lined up in the post-accident positions. Any leakage from the isolation valves in these systems is. determined methods and the results are added to the Type A test.

+ local Systems that operate in post-accident conditions filled with fluid as defined in Section 6.2.4 need not be vented or drained for the Type A test. Systems which form closed Seismic Category I systems inside Containment (as defined by GDC 57) are not vented to the containment atmosphere.

Leakage testing of instrumentation lines that penetrate Containment is done in conjunction with the Type A test. These lines will be. open to the containment atmosphere. Liner plate weld leak chase channels will not be vented during the Type A test.

All systems which are provided with isolation capabilities to satisfy GDC 55 or 56 are either. normally open to the containment atmosphere or are vented to the containment atmosphere during the Type A tests'able 6.2.4-1 contains the applicable GDC or other defined criteria for the isolation valve arrangements provided.

The electrical penetration pressurization system, supplied by dry pressurized nitrogen, serves to exclude moisture-laden air from each containment electrical penetration. During the Type A test, the nitrogen pressure in each electrical penetration will be locked in by shutting each penetration's nitrogen supply valve. Nitrogen supply to the penetration pressurization system will be isolated and the system headers vented to the outside atmosphere.

37 6.2.6-2 Amendment No. &

SHvVfPP FSAR type A test, the steam generator secondary side is to be vented

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SHNPP FSAR The service water lines to the emergency containment air coolers are neither vented or drained, as these lines are designed to GDC 57. The coolers may be required to cool the containment atmosphere during'the Type A test.

Pressurized gas and water systems are isolated downstream of the outside isolation valve Eor the system and vented outside of the Containment. This is done to preclude inLeakage into the Containment and to expose the outside isolation valve to an atmospheric back pressure to obtain accurate leakage characteristics.

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The reactor coolant drain tank, pressurizer relief tank, and the accumulator tanks are vented to the containment atmosphere. This is done to protect the tanks from the external pressure of the test and to preclude leakage to or from the tanks to help assure the accuracy of the test results.

The Eollowing systems are considered closed systems ins'ide containment that need not be vented and drained Eor a Type A test:

a. 4') Main Feedwater System 4 <) Auxiliary Feedwater System C Q Steam Generator BLowdown System II J y() Safety ReLated Portion of SW. System to and from emergency fan cooLers AH"1 through AH-4 g g) Portion oE component CooLing Mater System (to and Erom Reactor CooLant Drain Tank HX and Excess Letdown HX) 9g) Portion of the Steam Generator Sampling System Lnside Containment Out to the Containment Isolation Valve The system design meets the Eollowing requirements oE SRP 6.2.4.II.O Eor a closed system inside containmeht:

a) The system does not communicate with either the reactor coolant system or the containment atmosphere.

b) The system is protected against missiles and pipe whip.

c) The system is designated seismic category L'.

d) The system is classified Safety Class 2.

e) The system is designed to withstand temperature at least equaL to the containment design temperature.

E) The system is designed to withstand the external pressure from the containment structural acceptance test, g) The system is designed to withstand the Loss-of-coolant-accident transient and environment.

S7 6.2.6-3 Amendment No. ~

SHNPP FSAR 6.2.6.1.4 ILRT Test Method The air used to pressurize the Containment is conditioned for temperature and water vapor to prevent moisture condensation in the Containment at the test pressure. The air used to pressurize the Containment is essentially oil-free to prevent coating of the containment wall with oil or interfering with the test instrumentation.

Sensing devices are located at different locations in the Containment to measure average temperature and humidity. Location of the temperature and humidity sensor's are made with consideration to their respective patterns in the Containment. These patterns are employed in determination of the mean representative temperature and humidity for the absolute method of leakage rate testing. These data are periodically monitored during the test and analyzed as they are taken so that the leakage rate and its statistical significance is known as the test progresses..

The leakage rate test period extends to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of'ustained internal pressure. it If is demonstrated to the satisfaction of the NRC that the leakage rate can be accurately determined during a shorter test period, the agreed upon shorter period may be used.

'I At'he conclusion of the leakage rate test, the accuracy of the Type A test is verified by either of the supplemental test methods described in ANSI/ANS 56.8-1981, Appendix C. The supplemental test injects into or bleeds from the Containment an accurately measured amount of air. The supplemental test method selected is conducted for a sufficient duration to establish accurately

'he change in leakage rate between Type A test and the supplemental test. The difference between the supplemental test data and the Type A test data shall agree within 0.25 L Except as noted below, the following aspects of Type A testing follow 10CFR50, Appendix J guidelines are adhered to:

a) Pretest requirements including a general inspection b) Conduct of tests c) Acceptance criterion d) Periodic retest schedule e) Inspection and reporting of test If during the performance of a Type A test the leakage rate exceeds the criterion of 0.75 La , corrective action will be required.'f excessive leakage occurs through locally testable penetrations or isolation valves to the extent that it would interfere with the satisfactory completion of the test, these leakage paths will be isolated and the Type A test continued until completion. A local leakage test will be performed before and after the repair of each isolated leakage path. The calculated integrated leak rate will be obtained by adding the post-repair local leak rate to the measured integrated leak rate, and it will be required to fall within the acceptable limits of 0.75 La.

31 6.2.6-4 Amendment No. +C

SHNPP FSAR If any periodic Type A test fails to meet a&~~ 0.75 L the test schedule for subsequent Type A tests will be reviewed and approved by the Nuclear Regulatory Commission. If two consecutive Type A tests fail to meet

~her 0.75 L , a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet either 0.75 L At that time, the normal test schedule allowed by 10CFR50, Appendixa J shal be in effect.

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37 Amendment No.

6.2.6-4a

SHitPP FSAR For larger test volumes, a pressure decay method may be utilized to determine the leakage rate.

The total leakage rate Eor Type B and C tests will be less than 0.6 L . The individual testing performed on valves requiring a Type C test is described in Technical Specifications.

I In accordance with 10CFR50 Appendix J III.C.L, valves may be tested in the non-accident pressure direction when it can be determined that the results from the tests for the pressure applied in the non-accident direction will provide equivalent or more conservative results.

The criteria for determining the direction in which the test pressure is applied to the isolation valves is as follows:

a) Check, ball, plug, and non-werlge disc gate valves are tested in the accident pressure direction.

b) Wedge disc gate, butterfly, and diaphragm valves are tested in either direction since seat Leakage is the same in either direction.

c) Globe valves may be tested in the non-accident pressure direction if the test pressure would tend to unseat the valve and the accident pressure would tend to seat the valve. Where globe valves unbalance plug with flow the plug are installed such that the Elow and accident pressur e in the sam 'ction, the valve is tested in the non-accident ure direction. In 'ase the flow and the accident pre e will tend to seat the valve, while the n ccident pressure ' will to unseat the valve (i.e., force under the plug a cting a the actuator spring'orce). 37 Where globe. valves (with flow from the seat) are installed such that the flow and accident pressure not in the s irection, the valve is tested in the non-accident sure direction. In this ca he accident pressure wiLl tend to s the valve, while the flow and non-acci e ressure ' will tend to eat the valve. In both of these cases the test result zde equivalent or more conservative results.

6.2.6.4 Schedulin and Re ortin of Periodic Tests Types A, B, and C tests will 'be conducted at the intervals specified in Technical Specifications. These intervals are in accordance with Appendix J .

. to 10CFR50, with the exception of the testing of the air locks as described in Section 6;2.6.2.

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Perxo xc lea estxng o t e contax,nmen x o ation va ves nee ot be done during a refueling outage but may be scheduled at any time during an operating cycle. However, the test interval for any valve shall not exceed two years.

6.2.6-8 37 Amendment '.Vo. +

SHNPP FSAR h) To provide systera materials which are compatible with fluid cheraistry and applied codes and standards. System component design data parameters are given in Table 6.5.2-1.

6.5.2.2 S stem Design The system flow diagram is shown on Figure 6.2.2-1. System component design data parameters are given in Table 6.5.2-1.

A discussion of the spray header design inclurling a description of the nuiaber of nozzles per header, nozzle spacing, and nozzle is contained in Section 6.2.2.

System operation is automatically initiated by a HI-3 signal. The signal the two spray puraps and the motor operated spray isolation valves. 'tarts Within 55 seconds, water will reach the nozzles and start spraying (see Section 6.2.2). The motor operated NaOH isolation valves will be opened autoraatically by the HI-3 signal.

After the opening of the NaOH Isolation valve, the kinetic energy in the eductor will create a negative pressure to draw the Sodium Hydroxide solution (NaOH) frora the containment spray additive tank NaOH solution wi11 be injected into the Containment Spray System (CSS) lines just up stream of the CS pump suction at a rate sufficient to provide the required range of pH 8.5-il for the containraent spray. Turbulence in the fluid passing through the pum is 37 Additional NaOH can be added to the tank or through an eraergency NaOH addition line outside the Tank Building. IE necessary, the operator may reopen these NaOH isolation valves at any later time. The containment spray pumps initially take suction from the refueling water storage tank (RWST). The miniprum operating capacity of the RWST (see Section 6.2.2) is raore than adequate to supply enough water for the injection mode of operation. When low-low level tank water level is reached in the RWST, pump suction is txansferred to containment recir'culating sump autoraaticalLy by opening the recixculation line valves and closing the valves at the outlet of the RMST.

The Containment Spray Systera can provide one year of operation if required, I

The layout of the containment spray system headers and nozzle orientation (see Section 6.2.2) provides a minimum spray coverage of 92.6 percent of the containment free volume and 95 percent of the surface area of the operating floor (Elevation 286 ft.) with only one spray train in operation. This includes the volume beneath the grating in the operating floor. The specified grating has 80 percent free area. The drop size spectrura is discussed in Section 6.2.2.

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ESF ACTUATION SYSTB4S-SAFETY INJECTION SIGNAL(S)

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.E-10 (18-SB) Electric Equipment Protection Room Ventllatlon Exhaust Fan Stop Note 1 9, 4,5-1 CAR-2 I 66-8-430S H31,29 4J E-17 (IX-NHS) RAB Normal Exhaust Fnn ~ AN Stop 9. 4.3-2 CAR-2166-8-430S H31 ~ 35P I

Vl E-18 (IX-NNS) RAB Normal Exhaust Fan AKB Stop 9,4 3-2 , CAR"2166%-430S H31 ~ 35Q E-19 ((X&HS) RAB Normal Exhaust Fnn A&B Stop 9~4 3-2 CAR-2166%-430SH31 ~ 35R E-20 (IX-HHS) RAB Horme I Exhaust Fan AAB Stop 9 43-2 CAR-2'l66&-430$ H31 35$

E-6 (lh-SA) RAB Eeerqency Exhaust Fan h Start Note 1 9, 4,3-2 CAR-2166-8-4305 K31 33 E-6 (1B-SB) RAB Emerqency Exhaust Fnn 8 Start Note 1 9, 4,3-2 CAR-216&8-430S H31 ~ 33A S-3 (Ih<IHS) RAB.Homal Supply Fnn AIIB .Stop 9 4,3-2 CAR-2166%-430S H31,37 J S-3 (18-NHS) RAB Normal Supply Fan AN Stop 9,,4,3-2 CAR-2166-8-4305 H31 ~ 37J 1A-SA Emerqency Load Sequencer Panel A Start N/A CAR-2}66-G-50950~C IB-SB Emerqency Load Seauencer Panel 8 Start N/A CAR"216&G-509502 E-61 (IA-SA) Olese I Generator Bldq. Exhaust Fan A Start Note 1 9, 4,5-2 CAR-2166-8-430 E-61 (iC-SB), D I ese I Generator Bldq Exhaust Fan 8 Start Note 1 9,4,5-2 CAR-2 16&8-430 IS 2S- -I Cont, Atmos. Rnd. Honltor Isol ~ h -'se 2SP-V304A-1 Cont. Atmos nt h Close 2S P-V 302SB-1 Cont. htaa nt, Is 8 Close 2S P-V 303S B-l Atmos Sys, Cont, Isol ~ Close 2S- -1 Cont, Atmos, Sys. Cont, Isol ~ se

TABLE 7.3 ~ 1- (Cont lnued)

ESF ACTUATION SYSTEMS-CONTAIHMEHT ISOLATION PHASE-A (T)

EQU I PMEHT ACTUATION REFERENCE SCHEMAT I C/LOG I C I DEHT IF I GATI ON SERV I CE CHANNEL . ACTION FIGURE NUMBER NUMBER AH-37 (1A-NSS) Hon-Nuclear Safety Containment Fan Cooler AdB Stop 6,2,2-3 CAR-2166-8-430SH31,64 A AH-37 (18-HSS) Hon-Nuclear Safety Containment Fan Cooler AdB Stop 6,2,2-3 CAR-2166-8-430SH31,64A AH-38 (IA-NSS) Hon-Nuclear Safety Containment Fan Cooler Stop 6,2,2-3 CAR-2166-8"430SH31,64A AH-38 (lh-HSS) Hon-Nuclear Safety Containment Fan Cooler AdB Stop 6.2,2-3 CAR-2166-8-430SH31,64A AH"39 (18-NSS) Hon-Nuclear Safety Containment Fan Cooler Ads Stop 6 2,2-3 CAR"2166-8-430SH31,64A AH-39 (18-HSS) -'on-Huclear Safety Containment Fan.

Cooler 'AdB $ top 6,2,2-3 'AR-2166-8-430SH31,64 A 4J I

2FP-V44SA-I Fire Protection-Containment Fire I Hose Riser Isolation h Close 9 ~ 5 ~ I "4 M 2FP-V45SA-1 Fire Protection-Containment Mater Spr Ink I er I so I atl on A Close 9.5 '-4 2CT-V255A-1 Containment Spray Header Recirculation I so I etio'n A Close 6,2,2-1' CAR-2166%-423 2CT-V49SB-1 Containment Spray Header Recirculation I sol at I on 8 Close 6.2.2-1 CAR-2166-G-423 2CT-V8SA-I Containment Spray Eductor Test Valve Isolation A Close 6,2,2-1 CAR-2166-G-423 2CT"V145SB-1 Containment Spray Eductor Test Valve I sol ation 8 Close 6 2,2-1 CAR"'2166-G-423

+~oq s~l>

2SP-V300SA"I Conte I nment kfmJR. System A Cont, I so I at Ion Close 8

2SP-V308SB-1 Conte I nment kteoa. System% Cont, 0

I sol ation 8 Close 2SP-V309SB-I Containment @baca- System 4 Cont, Isolation Close Gv

>9P 'PVi)SA 3 C~~~~~T H> ~4~ Sys1?W /) ~J: Xpo4fio~ Zr pe- p'/FsA-a co4il~sa1I J/i, Std+ +JSw g &~~do7roH /) )

TABLE 7 3,>-7 (Continued)

ESF ACTUATION SVST84S~NTAIN4ENT ISOLATION PHASE-A (T)

EQUII)4ENT ACTUAT ION REF ERENCE SCHB4AT IC/LOG IC IOENT IF I CAT ION SERV ICE CHANNEL ACTION F IGURE NU4BER NU4BER 2SP-V301SA -g Cont. ~~5.

/Ig SiV pipe/y System A Cont, Iso, Close 37 V31$ 5B Q 2SP-W~ Contalnoent H2 Sampllnq System 8

't3)SD~ ~ Isolation Close N/A 2SP-VS4+I Conte'Inment H2 Sampling System 8 I so let Ion Close N/A 2AF -Y162S AB-1 Hydrazine to AFW Steam. Generator AIU) Close 2AF-V 163SAB-1 to AFW Steam Generator lh 1A'mmonia ALB Close 2AF-V)64AB-I Hydrazine to AFW Steam Generator 18 AH Close 2AF-V 165SAB" I Ammonia to AF'W Steam Generator 18 AIt,8 Close 2AF-V >66SAB" 1 Hydrazine to AFW Steam Generator 1C AI8 Close 2AF-V 167SAB-1 Ammonta to AFW Steam Generator 1C ALB Close 2SP-V 408SB-I PASS Isolation 8 Close 2SP-Y 409SA-I PASS Isolation A Close 2SP"V406SB-I PASS Isolation 8 Close 2SP-V407SA-1 PASS Isolation A Close 1A-SA Containment Spray Puop Interlock Trtp . Stop Note 1 6 2,2-1 1B-SB Containment Spray Pump Interlock Trip Stop Note 1 6,2 2-1 Notes; Interlock applicable only ~hen the reclrculatlon valve of Its respective safety train Is open, Ceaynium/Ass. //tau. Sysfrn A CksS codJ- Cso(i1iooi CksE C(osE CLosE

TABLE 9.1.3"2 (Continued)

FUEL POOL COOLING AND CLEANUP SYSTEM PARAMETERS Fuel Pool Cooling Pump Quantity (per FPCCS) 2 Type Horizontal Centrifugal Design flowrate, gpm 4500 %5l o TDH> ft. H20 m na Motor horsepower 150 Design pressure, psig 150 Design temperature, F 200 Material Stainless Steel Spent Fuel Pools Pool 1 Pool 2 Pool 3 Volume gals. 403,920 403i920 191,480 Boron concentration, ppm 2,000 2,000 2,000 Liner material Stainless Steel Stainless Steel Stainless Steel New Fuel Pool Volume, gals. 147 9 804 Boron concentration, ppm 2,000 Liner material Stainless Steel Fuel Pool Demineralizer Filter Quantity (per FPCCS) 1 Type Flushable Design pressure, psig 400 Design temperature, F 200 Flow, gpm 325 Maximum differential pressure across filter element at rated flow (clean filter), psi Maximum differential pressure across filter element prior to backflush, psi 60

SHNPP FSAR c

~ ~

crane prevents disengagement of a fuel assembly from the gripper during an 8SE.

The following safety features are provided for in the fuel transfer system.')

Transfer car permissive switch - The transfer car controls are located in the Fuel Handling Building; and conditions in the Containment are, therefore, not visible to the operator. The transfer car permissive switch allows a second operator in the Containment to exercise some control over car movement if conditions visible to him warrant such control.

Transfer car operation. is possible only when both lifting arms are in the down position as indicated by the limit switches. The permissive switch is a backup for the transfer car lifting arm interlock. Assuming the fuel container is in the upright position in the Containment and the lifting arm interlock circuit tails in the permissive condition', the operator in the Fuel Handling Building still cannot operate the car because of the permissive switch interlock. The interlock, therefore, can withstand a single failure.

b) Lifting arm (transfer car position) Two redundant interlocks allow lifting arm operation only when the transfer car is at the respective end of its travel and therefore can withstand a single failure.

Of the two redundant interlocks which allow lifting arm operation only when the transfer car is at the end of its travel, one interlock is a position limit switch in the control circuit, The backup interlock is a mechanical latch device on the lifting arm that is opened by the car moving into" posi tion.

c) Transfer car (valve open) An interlock on the transfer tube valve permits transfer car operation only'hen the transfer tube valve position switch indicates the valve is fully open.

d) Transfer car (lifting arm) - The transfer car lifting arm is primarily designed to protect the equipment from overload and possible damage if an attempt is made to move the car when the fuel upender is 'in the vertical position. This interlock is redundant and can withstand a single failure.

The basic interlock is a position limit switch in the control circuit. The backup interlock is a mechanical latch device that is opened by the weight of the fuel upender when in the horiiontal position.

e) Lifting arm (refueling machine) - The refueling canal lifting arm-is interlocked with the manipulator crane. Whenever the transfer car is located or the manipulator crane is over the core.~

f) Lifting arm (fuel handling machine) The lifting arm is interlocked with the spent fuel bridge crane. The li'fting arm cannot be ~eea~ unless the spent fuel bridge crane is not over the lifting arm area. IouJetcg the engaged gripper is in'he full up position or the disengaged gripper is withdrawn into the mast, 9.1 '-15

SHNPP FSAR In the unlikely event of a phase reversal prior to drive operations the crane

~

drives cannot operate by reverse phase relay action. In the event of a phase

~

reversal during hoist motor operation, the reverse phase relay will

~ ~

immediately operate to shut down the hoist drive, set the holding brake, and stop the load.

The crane is designed to maintain its structural integrity and hold its load under the dynamic loading conditions of the SSE. broad drop is precluded due to its redundant supporting system as described in Section 9.1.4.2.2.7 and Table 9.1.4-1.

9.1.4.4 Ins ection and Testin Re uirements As part of normal plant operations, the fuel-handling equipment i's inspected prior to the refueling operations. During the operational testing, procedures are followed to affirm the correct performance of the fuel handling system interlocks'he test .and inspection requirement for the equipment in the fuel handling system are:

a) Hanipulator crane, spent fuel bridge crane, rod cluster control changing fixture, and new fuel elevator.

lVI~BF I The minimum acceptablehtest shall include the following:

li Manipulator Crane and Spent Fuel Bridge Crane shall be load t'sated at 125 percent of the rated load.

2) The equipment shall be assembled and checked for proper functional and running operation.

l.l.W -(&a 37

~ I 1 i ~ Amendment No.

SHNPP FSAR

&ovc~c4 r o$ S~ea+ ~ueh luau'(Q.~g

+he (~el ~~gli~~

For condition (4) stresses do not exceed 90 percent of the elastic Limit oE the material.

c) During the construction phase of the project the use of the crane shall.

be controLLed to assure that the Life capacity and other operating limitations are not exceeded as required in the applicable s ecifications. Following completion of the construction use and prior to 37 applicable specifica 'o o assure compliance with performance requirements.

~CZurb<shed, d) Minimum operating temperature of the crane is 50 F.

e) All Eerritic material which is used in load bearing structuraL members are impact-tested to determine fracture'toughness of the material. Load bearing structuraL members are defined as structuraL members stressed in the process of transferring hook loads (vertical or horizontaL) through the crane to the main runway. ASTM A-514 material is not used in any load bearing structuraL members', other low alloy steel may be used with CP&L's (or it' agent's) written approval.

Either drop weight test per ASTM E-208 or Charpy tests per ASTM A-370 may be used for impact testing. The minimum operating temperature, as obtained by following procedures in Subarticle NC-2300 or ND-2300 oE ASME '8&PV code,Section III, Div. 1, based on the drop weight test or the Charpy V-notch impact test respectively, are not higher than 50 F.

E) Welding is performed by using welding procedures, velders, welding operators,'nd tackers qualified in accordance with AWS Dgl.l ~

g) Postweld heat treatment oE welded assemblies is perEormed, if necessary, when an assembly is under restraint during welding, when machining is to be performed, or Eor veLded steel greater than L-l/2 in. in thickness at the welded joint. Welds on al.l Load bearing structural. members are Postveld he@-treated in accordance with Subarticle NF-4620 of ASME B&PV Code,Section III, Div. 1, or other requirements as approved by CP&L (or its agent).

h) Where practical. weld joint designs susceptible to laminar tearing are ~

not used. Weld. joints susceptible to Laminar tearing are ultrasonical ly tested Eor soundness of base metal. and wel.d metal of the completed veld joint.

i) FuLL penetration butt weLds on aLl. Load-bearing structural members are 100 percent radiographed Eor soundness oE weld metal. and base metal where accessibLe. 'FuLl penetration tee welds on all. Load-bearing structural members and Eull penetration butt 'welds on aLL Load-bearing'structural members which cannot be radiographed are tested as follows.')

Magnetic particle or Liquid penetrant rest of root pass and final weld Layer.

2) ULtrasonic test of completed veld joint for soundness of veLd metaL and base metaL.

ALl fillet weLds and partial-penetration wel.ds are visuaLly inspected in accordance with and to the acceptance criteria of AWS Dl. 1 Paragraph 9.25.

FiLlet welds and partial-penetration vel.ds joining l.oad-bearing structural 9.1.4-21 Amendment No. ~

37

SHNPP FSAR Fire hose is hydrostatically tested in accordance with the recommendations of NFPA 1962, "Fire Hose Care, Use, Maintenance". Hose stored in outside hose houses will be tested annually. Interior standpipe hose will be tested every three years.

The standpipe system is designed and sized to provide, to the most remote hose

-station, the flow rate and pressure required for effective hose streams.

Operation of a hose station associated with a particular riser is al.armed local.ly and alarmed and annunciated at the Main Fire Detection Information Center (MFDIC) in the Pl.ant Communications Room and the Contro'1 Room foll.owing sensing of water flow in the standpipe riser by system flow switches.

Sectional shutoff val.ves provided for standpipes serving hose stations in safety related areas are located outside the safety related areas to permit access during a Eire.

Portions of the standpipe and hose systems installed in the Containment, Reactor Auxiliary'nd Fuel Handling Buildings, as shown on (Figures 9.5 1-2 and 9.5.1-4), are designed to be operable, if ~

needed, for manual Eire control in areas required for safe plant shutdown fol.lowing a safe shutdown earthquake (SSE). These portions of the standpipe system were analyzed for SSE loading and seismically supported to assure system pressure integrity. The piping and valves for these standpipes are designed to .satisfy ANSI B31.1, "Power Piping."

Normal.ly, the post-SSE standpipe hose station header is supplied from the Eire protection water distribution system through seismically qualified check valves. Following an SSE event, water supply for the post-SSE portion of the standpipe system can be obtained by local operator manuaL positioning oE val.ves to connect the Seismic Category I Emergency Service Water System, located in the Reactor Auxiliary Building, to the post-SSE hose standpipe header. Seismic Category I water supply is provided Eor the Post SSE Fire Protection Standpipe and Hose System by the emergency service water booster pumps. The ESW booster pumps are normally used following a LOCA to provide high head cooling water to. the containment Ean coolers. However, in the event of a Post-SSE fire, the ESW pump (A or B) and ESW booster pump would be started., This arrangement provides sufficient TDH to supply the two most remote hose stations with 75 gpm"(each) of ~ater at approximately 65 psig as discussed in NFPA. The seismic check valves prevent outflow to other portions of the fire protection water distribution system, which may have failed during

. the seismic event, and thus avoid loss of hose line protection after the earthquake.

c) Self-Contained Breathin E ui ment - Breathing equipment is provided as required for protection against smoke inhalation of personnel required to be in plant areas to control.. fires or to continue vital plant operations.

Self-contained breathing apparatus, using full. face positive pressure masks, approved by National Institute for Occupational Safety and Heal.th (NIOSH),

with a.minimum capacity of one half hour, are provided for fire brigade and control room personnel.

Ai

<<~" c+

ikiO~c,l 1&

%oe=aecea- air"bottles ~15 located onsite for each self-contained breathing unit, used by Eire brigade and control room personnel, with an onsite six hour 37 Amendment No. ~37 9.5.1-25

SHNPP FSAR 4epwee~ 195 F ~cl ~ ~>~

During periods of diesel generator standby, the jacket water cooling system

~

is automatically maintained ~h5!h=B by means of an electric jacket water [37 keep-warm heater, jacket water keep-warm thermostat, and a motor driven jacket

~

water keep"warm pump (see Figure 9,5.5-1). High and Low temperature alarms monitor the jacket water temperature and the "keep warm" pump is tripped automatically upon start of the engine. The jacket water heater is provided with power from a non-safety power distribution panel in the diesel enerator building.

gs<<~ soho~~ ~a<48~0e Engine cooling water system design precludes trapping of air wet xn t e engine spaces. Vents are pr'ovided in the jacket. water cooling system standpipe in order to assure that all spaces are filled with water. Provisions are provided to treat the jacket water by adding or removing chemicals. Corrosion and organic fouling are controlled by and or other suitable chemicals and biocides. The chemicals utilized are compatible with the system materials and each other. The pH of the jacket cooling water is maintained between 8.25 and 9.75 as specified by the manufacturer.

~(~go m q5 F 1

The jacket water heater is conservatively sized and will maxntaan packet water

~ I7OC 37 A description of the Diesel Generator Building Ventilation System is provided in FSAR Section 9.4.5.

total heat rejection at 110 percent load from the jacket water heat

~

The exchanger is 18,078,456 Btu/hr based on 95 F Emergency Service Mater maximum

~

inlet temperature. The heat exchanger is designed to a duty of 20,662,000 Btu/hr.

9.5.5.3 Safet Evaluation The Diesel Generator Cooling Water System is designed to have adequate capability to carry away the waste heat from diesel generator units under all loading and ambient conditions. The diesel generator is capable of operating fully loaded without secondary cooling for a minimum of one minute. Sufficient water is contained-in the engine and standpipe to absorb the heat generated during this period. The normal supply of cooling water Eor the diesel generator is the normal service water pump. Upon loss of offsite power"the emergency service water "pump will supply cooling water to the diesel generator aEter a period of 20-25 seconds.

The Diesel Generator vendor, Transamerica De Laval, ran a continuous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Load test on a diesel engine-generator set similar to Shearon Harris'nit.

The test engine ran for 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> at 100 percent Load, followed by two (2) hours at 110 percent load. Test indicated that less than'three (3) gallons of water was Lost due to evaporation, boil off, and minor Leaks.

S7 9.5.5-2 Amendment No. +

ice, PC.J Q Lg-ID'- 2 l4>@Are SHNPP FSAR

~y,per M~g g Q~ ~~uual udbi4r'oN cd- polo'Ore ~o her.

The NPSH requirement for the engine jacket water pump orrespon s to a minimum standpipe level oi 53 1/16 i.n. Normal wacer leveL is Assuming a standpipe water level of 186 1/2 in. at start of seven (7) days of continuous 100 percent load operation, approximately 400 gallons of water is available between elevation 186 1/2 in.

and 53 1/16 in. The standpipe, wi,th a capacity of 400 gallons, provides raore than adequate water to maintain the requirerl pump NPSH and make-up for seven days of continuous operation.

r AL1 components of the Diesel Generator Cooling Water System are designed co Seismic Category I requirements. The jacket water heat exchanger and connections to the Emergency Service Water System are also designed to Safety Class 3 requireraents ~ Failure of any non-Seismi.c Category I structures and coraponents vill not affect the safety related performance of the system. The diesel engine-mounted cooling water system piping and components, meet the guidelines as stated by the DEMA standards. The design stresses which includes mechanical, pressure, thermal, and seismic induced loads for the engine-mounted piping have been determi.ned by the Di.esel Generator manufacturer (i.e., Transarnerica DeLaval'ncorporated (TOZ)) to be well withi.n allowable stresses as permitted by ANSI 831.1.

py'he The TDI approved QA/QC program used i.n conjunction wi.th the manufacture of diesel engi.nes, engine-mounted components, and piping comply with the requireraents of Appendix 8 of 10CFR50.

Each diesel generator has its heat exchanger's tube side connected to the respective emergency service water system train.. Therefore, a si.ngle failure of a component, or the loss of a cooling source 'will not reduce the safety related functional performance capabi.lities of the system. The jacket water standpipe is provided with low level instrumentation for leak detection. In addition each diesel genera'cor room is equipped wi.th a sump and sump puiap to collect and dispose of leaking fluids within the Di.esel Generator Bui.lding. The sumps are provided wi.th safety class 1E level Lnstruraentation with annunciation in the Control Room to alert operators of potential flooding. The sump pumps are automati.cally actuated on high sump level.

This system is housed in a Seismic Category I Structure (Diesel-Generator Bui.lding) thac is capable of withstanding the effects of natural phenoraena such as earthquakes, tornadoes, hurricanes, floods, and mi.ssi.les. As shown on Figures 1.2.2-86 and 1.2.2-87, each .di.esel generator is located 4n a separate roora. The protection of safety-relaterl systems frora the effects of hi.gh and moderate energy pipi.ng failures are considered Ln the design of the diesel genr racor facility. The facili.ty does noc contaLn any high energy lines but does contaLn the moderate energy lines of the Eraergency Air, Fire Protection, Fuel Oil, Lube Oil, Miscellaneous Drains, Service Water, Potable Water, and Station Air systems. Facility design provides for the effects of failures (cracks) in these moderate energy fluLd systems'looding from cooling line leaks does not irapact other diesel generator areas where the line break has not been postulated. Facility design as shown on Figures 1.2.2-86 and 1.2.2-87, and sump drain design as shown on 1'igure 9.5.5-2 precludes flooding impact on the unaf fected diesel generator area. A further discussion of the postulated piping failures in high and moderate energy fluid systems is ti loca ted i.n Sec on 3. 6.

57 Amendment No. S(

9.5.5-3

SHNPP FSAR 9.5.5.4 Testin and Ins ection The diesel generator jacket system will be tested during the periodic diesel generator tests as described in Section 8.3.1.1, and its standby (keep warm) condition shall be raonitored per plant operating procedures'ystem instrumentation loop checks and setpoint calibration will be performed in accordance wi.th the SHNPP Preventive Maintenance Program or at each refueling outage, vhichever is the earliest.

All raaintenance on the eraergency diesel generator vill be followed by a verified line-'up and post-maintenance test in accordance with the surveillance requireraents of Technical Specifications. The line-up procedure will verify that the keep~arm system is properly aligned. Testing of the diesel generator operation does not require realignment of this keep-warm system.

The cooling water in the closed loop system is periodically analyzed to monitor its condition and treated as requi.red to raaintain its quality.

9.5.5.5 Instrumentation A lication The following alarra points with local annunciation are provided in the Diesel Generator Cooling Water System for each diesel generator:

a) jacket vater inlet high/low teraperature b) jacket water outlet high/low temperature

.) jacket vater high temperature trip d) standpipe low level e) jackee vater pressure f) jacket water low pressure trip Jacket vaeer pressure swi.tch (PS-22C) and jacket water low pressure trip (PS-21C) are separate pressure switches vhich are connected to a comraon process tap (refer to Figuxe 9.5.5-1).

Pressure settings for jacket water pressure s~itch is 12'si and decreasing (alarm point), jacket water low pressure trip switch is 10 psi and decreasing.

Operation of any of the above mentioned local alarms is indicated by annunci,ation on the Diesel Generator'ontrol Panel and also "trip" or "trouble" alarms on the Hain Control Board. In addieion, pressure and temperature devices are provided for local indication and thermocouples are provided for reraote indicati.on of temperature. Teraperature settings for jacket water low temperature inlet/outlet alarm svitch actuation is 140 F decreasing respectively.

In addition, tempexature settings for the jacket 37 waeer high temperature inlet/outlet alarm switch actuation is 175 F increasing and 190 F increasing respectively. One thermocouple is placed in the piping between the return header and standpi.pe (jacket vater outlet high/low temperature) and the second thermocouple is placed in the piping between the 37 9.5.5-3a Amendment No. M

SHNPP FSAR described in Section 11.5.2.6.5. The monitor is powered by the Emergency A Bus. A containment isol.ation actuation signaL will isolate this monitor from the Containment.

This monitor provides a radiation alarm when concentrations reach preset limits. The receipt of this alarm will alert the opera'tor to the presence of low level leakage so that additional sampling can be effected in order to locate the jeakage source. An interlock ~~be provided to terminat;e 37 continuous purge operation on high radiation. /4 12.3.4.2.8.2 Control Room Normal Outside Air Intake The control room normal outside air intake'lenum has two, beta sensitive monitors, one associated with A Bus, and one with B Bus. These monitors are part of the safety related portion of the RMS (Section 11.5.2.3) and use the ambient gas monitors described in Section 11.5.2.6.1.

These monitors provide a high radiation alarm when concentration Level.s reach preset limits. Upon receipt of the alarm; the monitor closes the normal

~

outside air intake valves associated with a given unit, stops the exhaust fans, closes the exhaust dampers, starts up the emergency filtration fans and opens the required valves and dampers to put the air flow into the recirculatory mode. The receipt of these alarms will also alert the operator to check the radiation Levels at both emergency outside air intakes, and to open the intake at which the radiation level. is lower (Section 12.3.4.2.8.3),

12.3.4.2.8.3 Control Rooms Emergency Outside Air Intake There are two emergency outside air intakes. Each intake has two duct-mounted Beta monitors associated with A Bus (Intake 10 and llA) and B Bus (Intake 10 and llA)~ These monitors are part of the safety related portion of the RMS (Section 11.5.2.3), and use the ambient gas monitors described in Section 11.5.2.6.1.

These monitors provide indication to the control room personnel of the radioactivity levels at each emergency air intake, thereby allowing the operator to choose which emergency intake to open (see discussion in Section 7.3.1.5.7 and 12.3.4.2.8.2). These monitors also provide a high radioactivity al.arm when concentration levels reach preset limits.

There is one outside air intake monitor for the Technical Support Center (TSC) ~ This monitor is also an in-duct ambient beta monitor. The radiation levels and any alarms are received on the RMS consoles in the Control. Room, the WPB Control Room, and the access control point. They are also monitored by the Emergency Response Facility Information System (ERFIS).

QQ. /-8* '37

/, n~ Amendment No. W

SHNPP FSAR 13.0 CONDUCT OF OPERATIONS 13.1 ORGANIZATIONAL STRUCTURE OF APPLICANT 13.1.1 MANAGEMENT AND TECHNICAL SUPPORT ORGANIZATION 13.1.1.1 Or anizational Arran ements Since the first nuclear generating unit belonging to CP&L began commercial operation in March 1971, the amount of nuclear generating capacity on the Company's system derived from nuclear power has increased substantially.

Accordingly, the Company's responsibilities in connection with its nuclear facilities have grown. During this period of time, the Company has developed and enhanced its capabilities with respect to the construction, operation, and maintenance of its nuclear facilities. The Company has safely managed H. B.

Robinson Unit 2, and Brunswick Units 1 and 2 since they were placed into operation. The Company has also managed the construe'tion of the Brunswick and Harris facilities'he Company has been, and will continue to be totally committed to safety and quality in the construction and operation of our nuclear facilities.

The Company has reorganized its management structure several times to accommodate and better manage the increased nuclear capac'ity and additional associated personnel. The most recent maj or reorganization, announced on September 1, 1983, reflects the strengths developed and lessons learned from the Company's operating experience as well as from the experiences of the rest

~

of the nuclear utility industry. It focuses the authority and responsibility for'peration, engineering, and construction under one individual at each of CP&L's three nuclear plant sites. In addition, it ties many of the related offsite nuclear support organizations to the Shearon Harris Nuclear Power Plant (SHNPP) and H. B. Robinson Steam Electric Plant (HBR) plant organizations and places them under one individual, the Senior Vice President Nuclear Generation. The Vice President, Brunswick Nuclear Project (BNP), who presently reports directly to the Senior Executive Vice President, Power Supply & Engineering and Construction, also benefits from the support services that are under the Senior Vice President - Nuclear Generation (see Figure 13.1. 1-1) . ~

The Company's nuclear projects are supported by an extensive organization that provides expertise in a variety of areas. For the most part, the organizations are structured to focus nuclear activities within separate departmental and organizational structures. This philosophy ensures that the Company's other, nonnuclear activities will not divert appropriate management attention from the conduct of its nuclear activities.

The Corporate support for nuclear activities is managed by the Senior Executive Vice President - Power Supply and Engineering & Construction Groups who reports to the President/ChairmanlM~e~ Executive. OE+ i'cer .

Reporting to the Senior Executive Vice President - Power Supply and Engineering & Construction are: a) Senior Vice President - Nuclear Generation Group'b) Senior Vice President Fossil Generation and Power Transmission Group> c) Senior Vice President - Operations Support Group,'3.1.1-1 Amendment No. g 37

SHNPP FSAR d) Vice President Brunswick Nuclear Project Department; and e) Manager Corporate Quality Assurance Department (see Figure 13.1.1-2). The responsibilities of each of these groups and departments are described below'.

13.1.1.1.1 Nuclear Generation Group The Nuclear Generation Group is responsible for providing offsite technical and managerial resources to assist and support the operating nuclear plants in areas of nuclear licensing, civil design, instrumentation and controls, computers, mechanical, electrical, nuclear engineering, metallurgical analysis, construction, operations, and industrial security.

Vice P~s,iden.<

Reporting to the Senior Vice President ",Nuclear Generation Group are'. 1) the Vice President - Harris Nuclear Project Department; 2) the Nuclear Project Department; 3) the Vice President -

k~~ " Robinson Nuclear Engineering 6 Licensing Department; 4) the Manager Nuclear Plant Construction Section', and

5) the Manager - Nuclear Staff Support Section (see Figure 13.1.1-3).

1) The Harris Nuclear Pro'ect De artment is responsible for managing the site activities in a'anner which will promote the economic, safe, reliable, and effective operations of the plant over its lifetime. The organization, formed on September more direct on-site management control over ~

1, 1983, represents the Company's concept of providing engineering, construction, startup, and operations activities at the plant. Other support functions are provided from other departments in Power Supply and Engineering 6 Construction.

The Vice President Harris Nuclear Project is responsible for managing ~

the Harris Nuclear Project. He conducts these activities in a manner which protects the health and safety of the public, is in compliance with the applicable governmental regulations, and is within the policies and guidelines of the Company.

Oper~4i o~

Reporting to the Vice President - Harris Nuclear Project Department are:

a) the General Manager Harris Plant Section,') General Manager Milestone Completion Section', c) General Manager - Harris Plant Engineering Section',

d) Manager - Project Administration Section; e) Manager - Planning and Controls Section', and f) Manager Completion Assurance Section (see Figure 13. 1. 1-4) .

< Oper~di~w a) The Harris PlantkSection is responsible for the startup testing, operati'on, maintenance, security, environmental and radiological control, and management of the plant in accordance with its construction permit or operating license, Technical Specifications and Plant Operating Manual (see Section 13.1.2).

b) The Milestone Completion Section is responsible for achieving the completion of project milestones in conformance with NRC regulations, code/,'rocedures, permits, specifications and drawings; and company policies and commitments'he General Manager Milestone Completion has direct management responsibility for plant construction and task responsibility for portions of engineering, startup, planning, and scheduling to ensure completion on schedule, within budget, and 'in 13.1 ~ 1-2 Amendment No. 2f 3'p

SHNPP FSAR

' strict c) compliance with commitments to QA and ALARA requirements.

The Harris Plant Engineering Section (HPES) is responsible for providing as required, detailed engineering modifications and maintaining design and procurement documents. The Section is supported by Nuclear Engineering and Licensing and/or out'side consultants'he Section provides engineering support for the construction and/or operation implementation of these modifications', to the operations organization in areas such as spare parts, Q"list equipment, and equipment qualification; and in the review of plant operating, maintenance and surveillance procedures as requested. A major benefit of this process is that the same technical staff that administered the design of the Harris Plant during .its construction will be responsible for the engineering support of plant operations.

d) The Project Administration Section supports the administrative needs of the Harris Nuclear Project Department by providing a centralized source for these services. The Section provides these services either through its own central organization location or through satellite offices located with the various organizations it supports. These activities span a range of responsibilities from coordination of some activities, such as training, employee relations and computer services coordination, to management responsibility for activities such as document control and warehousing.

' e) a The Planning and Controls Section aids management in ensuring that consistent, coordinated structure of work activities is achieved which focuses on the objectives and goals of the Department. The Section monitors the resulting structure and reports information to other site management indicating compliance with or variances from the plan. Primary responsibilities of the Section are to identify, develop, and implement programs, systems, methods, and related documents for planning and scheduling, budgeting, cost control, cost assurance, and industrial engineering such that management visibility is maintained to historical accomplishments as well as anticipated variances. Information and forward visibility permits corrective action while managerial alternatives remain open.

f) The Completion Assurance Section is responsible for handling special assignments as directed by the Vice President Harris Nuclear Project.

2) The Robinson Nuclear Pro ect De artment operates and maintains the Company s nuclear generating facility at the H. B. Robinson Plant.

&e responsibilities are similar to those &~ther

~ ~ ~ ~ ~

of C4 V"ce~aietcn+-Harris Aluc.(eg.r R-o~eH of W4 V'c.a &e,side a+-Rck ~KNuchxv BoOJ ~

3) The Nuclear En ineerin 6 Licensin De artment is responsible for the licensing and engineering support of the Company s nuclear generating facilities. Reporting to the Vice President Nuclear Engineering 6 Licensing 13.1.1-3 'Amendment No. 2C Z7

SHNPP FSAR (see Fi'guM t3 I I-4,) cx.&

DepartmentLare: a) the Manager Nuclear Licensing Section; b) the Managerg-c t e Manager Nuclear Engineering Projects Section.

ikd:lea.r Pin~< E<g L n.eeri ~ Sec%i'on.

The Nuclear Licensing Section acts as the Company's interface with a) the NRC Office of Nuclear Reactor Regulation and, for multiple plant activities, the Office of Inspection and Enforcement. The section is organized into four units with the following functional responsibilities'.

The Project Nuclear Licensing Units are responsible for coordination of Office of Nuclear Reactor Regulation (ONRR) activities affecting the Company's three nuclear projects.. This includes the coordination and preparation of responses to ONRR requests, and the preparation of license amendments and li.censing documents such as the Harris Final Safety Analysis Report (FSAR). These units are responsible for the maintenance of operating licenses, revisions .to the technical O~Cnik&idw ~

specifications, and updating of FSARs 74- urd~~ parfici'~Ve. in.indy~~~

Ctfi left OlVACll'~~ Qf dVP4.

~

The Special Nuclear Programs Unit is responsible for coordination of generic licensing issues'his includes coordination and preparation of responses concerning generic ONRR activities affecting the Company's four nuclear units. It advises Company management on critical licensing issues and ensures that incoming NRC correspondence is routed properly and that responses are prepared to address licensing issues accurately. In addition, Special Nuclear'rograms coordinates the Ace<'iraq Company's regulatory related involvement in industry organizations including AIF, EEI, and EPRI. This Unit also participates in various utility owners'roups and supports other special projects of E~Q a technical or regulatory nature as required. i Sc b) hb !ca Pl CP.

The[Engineering Section/

~

for providing engineering support for the Company's nuclear plants and

~ l5 uc.le&i P Ja.n+Me p responsible for utilizing feedback received from the operatin plants so as to prevent identified problems from recurring. The objectives are to provide engineering and procurement of engineered products on 'one'hem schedule with designs that are economical; safe, efficient, reliable, and compatible with the environment.

They provide the design engineering necessary operating plant feedback projectsa and to TKes4 Onxts are resolve those for identifying potential

~<

operating plant problems which staffed with engineers designers of required experience, education, and capability.

and e

Architect/Engineers)sin5 other consultants may also be retained to Znze.r+ R Reporfi+ 0 froze tLe W~+i

~

assist the Sections in meeting their objectives.

I'3.L I - 5

- Muc(nzl F'lax~FngirLeeri~ Sec Wi~Na a Wll R ')

~ t~ gaa~g6'rg uppoY'4> Qoc.(co.r Plaat @cgidng X a.~ zz '.718 4 Di f ec+-Y M4lc.le Safe% Aevi'e~ Uuf. 13. 1. 1<<4 en m

SHNPP FSAR c) The Nuclear Engineering Projects Section is divided into three units: Nuclear Pzojects Unit I, Nuclear Projects Unit II, and Engineering Administrative Unit. The Section is responsible, through its Nuclear Projects Units, for ensuring that the NELD provides the required design and engineering support for each nuclear project and that the nuclear projects appropriately utilize the resources of NELD. The nature of this support is reflected in defined written agreements with each of the projects and in accordance with other departmental procedures and/or guidelines. The Section establishes the "..

scope~ content, and magnitude of projects assigned to ~/><

and manages the A/E engineering work throughout the final acceptability of the design project.

The Engineering Administrative Unit provides the +eehnkr~ support services 'required by the Sections in the Department. Priorities are set to meet the identified schedules established for the nuclear projects. The Unit serves as the focal point for collecting, management to monitor schedule and cost progress on ~

processing, and disseminating required information to allow responsible assigned plant modification projects and provides support in engineering sched'ule preparation, engineering, scheduling services during project implementation, supplement scope development, QA records support, and other engineering administrative support to in-house engineering design sections within the Department; I

Zmest 8 The Director - Safety Review, Nuclear Engineering is responsible e <o <Le for reviewing documents generated by the Company's nuclear organization am of and A/Es identifying problems in engineered safeguards systems and cvrrad par}e plant safety features', assessing activities and trends in the industry A.l, I - ~ regarding design and operation of safety features', providing feedback to preclude potential nuclear safety problems in ongoing plant designs and design of modifications, and assuring that ALARA concepts for

'radiation contxol are considered in engineered designs.

4) The Nuclear Plant Construction Section manages the procurement and contracting activities required to support the completion of construction project assignments. The Section provides both firm-price and reimbursable contracts, onsite procurement and expediting services, and construction equipment and tool management. Onsite procurement staffs have been established at the Harris, Robinson, and Brunswick Nuclear Projects'he Section also provides support services .to the other Departments within the Company in the areas of estimating, budgeting, cost control, cost reporting, construction accounting, information management, and construction security.

5) The Nuclear Staff Su ort Section is primarily responsible for coordinating the implementation and maintenance of operationally oriented programs that require high technical knowledge of methods and procedures and that should be relatively consistent among the plants. The Section is also responsible for preparing reports and documents, performing staff studies, providing administrative/technical support as required and coordinating the Company's involvement in Institute of Nuclear Power Operations (INPO). These efforts are coordinated with each project.

~

pl ovlc{< o~~ccf ga~~ deco~)$ ~

NQP Pw 13 '.1-5 Amendment No. Pf 37

SHNPP FSAR 13.1.1.1e2 Fossil Generation and Power Transmission Croup The Senior Vice President Fossil Generation and Power Transmission Group is responsible for managing the Company's fossil and hydro generating facilities and the Company's transmission line facilities necessary to meet its bulk power requirements.

13.1.1 1.3

~

The Senior Operations Support Croup

~ Ic,o.l Vice President - Operations Support Group is responsible for the management of the materials and fuel needs of the generating and transmission facilities in addition to the training and technical support of those personnel.

Reporting to the Senior Vice President " Operations Support Grouplare: ~the-(~ Figuea IZ.l.l- /)

'Z) the Manager Materials Management Department,

4) the Vice President Nuclear Safety and Environmental Services Department Their responsibilities are summarized'elow! ..

Operations Training Department, and~4) the Manager-d"ndTcchm'cn.(~<v<'<c~

The Nuclear Fuel Section xs staffed wat personne avxng bot t e tec nical and managerial expertise X~r< B required to ensure a timely and adequate supply of nuclear fuel, to review fmove to fuel and core design, to support nuclear plant outages (including refuelings) page I X l.l-?)

and operations, and to provide for spent fuel management. The Nuclear Fuel Section meets with members of the Company's operating nuclear plants on a continuin basis to lan and o timize the fuel o eration strate lg) The Materials Mana ement De artment is responsible for corporate purchasing, inventory control, warehousing, and salvage of the Company's material needs

<) The Ope~f~wTra tn.i~pa& ~~M>~~'Odd par tme~+ su H~ ruPc.(~p'~z,'(

dreiy lsd raAnRcpssnllEod noc7,@cled 'Eve~~ 'Cdm /la.n':4 ciaso adrs ksa dl phts cs pter'ss mnsd prsvidcn rmdi lrdd; caotenvlronnsen alaoptoert RePortind to the Vice President POerati osnrTai ,in gnDlePar mte a~teer: (a) the g,~@c Nuclear Training S"ction, CMy Cd) fFa~e<

. 4) ~et PaQs'c

-Nuclea~ Fuel Sec<sdhpi ~ (c) gd-~~dt~~ C ~+~~Heal)Jpg <<~

%ton.ave CLQnn.ica.(Suppor $ ,

A+C Pgssre l3,t,( ldp) a) The Nuclear Training, Section provides support to the Nuclear Project Departments in the areas of Operations, Technical and Craft Training, and the operation of the simulators and other training 13.1.1-6 Amendment No. PK >7

SHNPP FSAR fhcilities at the HENDEC and at the respective nuclear projects. The primary purpose of the Nuclear Training Section is to assure that the Company has highly qualified personnel available to maintain and operate its nuclear generating plants in a safe and efficient mannex'.

b) Z~W 6 - Er~m

4) The Emergency Preparedness Section is responsible for: directing and compliance', asse'ssing the readiness of ~

coordinating Corporate Emergency Planning to ensure regulatory CP&L emergency plans and programs'serving as interface with regulatory agencies on emergency preparedness matters'providing emergency preparedness support for CPSL nuclear plants; maintaining training qualifications of plant personnel in emergency response', testing emergency preparedness by preparing and conducting exercises> ensuring the availability and operational and providing coordination with federal, state, and local agencies.

The Physical Sciences Unit within the Emergency Preparedness Section is responsible to provide'meteorological support to the Company's nucleax CIA f6 and fossil plants and other departments within the Company. .The primary purpose of the /nit is to provide meteorological data,

~c lz.l.l-l) forecasts and expertise to the Company's nuclear facilities, assisting in the dose projection capabilities of the plants'mergency response personnel by providing micro-meteorological forecasts and dispersion expertise. The 3nit also provides on a routine basis, early notification to appx'opriate Company personnel those sevex'e weather events which may affect either the generation of power or customer services.

~ ' C

4) The Nuclear Safet and Environmental Services De artment monitors and reviews~ plant nuclear safety conducts environmental assessments', performs chemical and materials laboratory services',Land provides staff support in. several technical disciplines.

('Sec F Is~re I C.l.l- II) 4lrecVa CLa emagenc~ p~~r~g~~ >r+~~ ~

Reporting to the Manager~ Nuclear Safety and Environmental Services ~

Department are: a) thetCorln'r~ate Nucleax Safety Section, b) thelG~M PrePrei and C8) the)Environmental Services Section, a) The Corporate Nuclear Safety (CNS) Section combines technical

'expertise in an integrated off-site/on-site program to monitor and evaluate plant nuclear safety performance. The organization fulfills requirements specified in ANSI N18.7 for Independent Review and NUREG-0737 for the Independent Safety Engineering Group (ISEG). In 2~c D- +rom Pa.~ IQ.l,l-p Q) +hxc-Y+ E fpo~ N lZolol~g Amendment No. P(37

SHNPP FSAR addition, CNS provides the base for the operating experience feedback

+MH program within CPRL as well as input to the establishment of priorities of nuclear safet~elated items.

The CNS independent review activity addresses the following'.

(1) Procedures and changes meeting 10 CFR 50.59 review criteria, (2) Licensing actions, (3) Test or experiments not described in the facility FSAR, (4) Plant operational occurrences (LERs),

(5) Regulatory violations (IE Reports),

(6) Technical Specification changes (7) Plant Nuclear Safety Committee (PNSC) meeting minutes, and (8) Any item deeme'd appropriate for review relative to safe operations.

N)d The Corporate Health Physics Section consists M personnel with education and/or work experience in fields of ~M~ hygiene or health physics. The section is also responsible for formulating and recommending corporate level health physics policies and programs, evaluating health physics programs and recommending any needed improvements and modifications in those programs, and providing health physics expertise throughout the Company. 'he Section provides support to the licensing and corporate nuclear safety activities of the g y Company, is responsible for the development and distribution of the Corporate ALARA Program, and makes periodic assessments of various gK ALARA programs developed to comply with the Corporate ALARA Program.

This Iection also conducts ayffannual audit of the QA program.

4) The Radiological and Chemical Support Section (RRCSS) provides staff support in the areas of health physics, chemistry, and radiological environmental activities and for the effective operation of the environmental, dosimetry, and chemistry laboratories. The RSCSS has responsibilities identified in the Corporate Emergency Plan to provide health physics and environmental support to the nuclear plants in the event of an accident. These responsibilities and services are

. the Health Physics Unit, the Environmental Unit and the Chemistr Unit.

~g) The Environmental Services Section +EEH-3-conducts the Company's environmental monitoring assessments and performs analytical chemistry and metallurgical laboratory services at the Harris Energy h Environmental Center (HENDEC) in New Hill, North Carolina. The Analytical Chemistry, Air Quality, Biology, and Metallurgy Laboratories provide an array of services and technical support to generating plants, engineering activities, quality assurance and construction 13.1.1-8 Amendment No. A 37

SHNPP FSAR programs within the Company.

The Water Environmental Regulations and Permits Unit is responsible for obtaining the National Pollutant Discharge Elimination System (NPDES) permits and together with the Air Environmental Regulation and Permits Unit, any federal, state, and local permits not required by the NRC.

c) X~avfd - 5m~ p~a lZ.l. I-7 13.1.1.1.4 Brunswick Nuclear Project Department The Vice President Brunswick Nuclear Project Department reports to the Senior Executive Vice Pxesident - Power Supply 6 Engineering and Construction. His responsibilities are similar to those of the Vice President Harris Nuclear Project Department, 13.1.1.1.5 Corporate Quality Assurance Department The Manager of the Corporate Quality Assurance Department reports to the Senior Executive Vice Px'esident - Power Supply and Engineering 6 Construction (see Figure 13.1.1-1) and is responsible for the quality assurance, quality control, and audit functions which were at one time performed separately for engineering and construction, operations, and corporate quality assux'ance audit activities'n this manner, the Manager - 'Corporate Quality Assurance oversees the QA/QC activities of the Power Supply and the Engineering 6 Construction Groups while maintaining independence from any responsibilities within those organizations Re&r<c ~<>@Seed~~~ lC.B fat' desati'p+ioN crF 0~

~

Qv~'p Accu~a.nn Ol niectM<~w.

orting to the Manager Corporate Quality Assurance Department dre: 1)

Harrx Plant QA/QC Section, 2) the Operations QA/QC Section, and 3) the Services ction (see Figure 13.1 1-13) ~ Their responsibilities ar

~

summarized b a) The Haxris P QA/QC Section has the primary r 'nsibility for the Harris Plant Quality Ass nce/Quality Contxol in t engineering and construction phase. Its pur e is to anticipa and preclude safety-related nonconformances. This section x iso resp able for the pxeparation of the ASME "N" Stamp QA Manual.

b) The Operations QA'/QC Sec is respo ible for assuring proper application of quality stan/a s, practices, an rocedures associated with plant startup, operatio ~maintenance or modificatio t CPGL operating plants.

ervices Section is responsible fox supporting CP 's nuclear c) plants 'he The areas of QA Engineering, vendor qualification/ survei for nce and tra 'ng. This section is also x'esponsible conducting an independen r orate audit ro ram.

-13.1.1-9 Amendment No X ZY

SHNPP FSAR Carolina Power & Light Company's management and technical support staff positions performing key functions for the SHNPP project are filled by individuals with several years of experience as presented in Table 13.1.1-1.

Corresponding resumes are provided in Tables 13.1.1-2"through 13.1;1-6.

The General Manager, Harris Plant Engineering Section is the "Engineer-in-Charge" as specified in ANS 3.1, September 79 Draft.

13.4.1-10 Amendment No. & Z7

TABLE 13 ~ 1 ~ 1-1 EDUCATION AND EXPERIENCE SR+SR IES FOR KEY PERSONNEL SUPPORTING SHNPP (as of 1985)

APPL I CABLE EXPERIENCE SECTION OR ORGANIZATION TITLE EDUCATION (YEARS )

NUCLEAR GENERATION GROUP r Pr ect Rll ~ Senior Vice President Nuclear Generation 'Group H, A, HcDuf fle BS Civil Eng. 37 Vlcc Presldent- R, A, Watson BS Nuclear Eng, 29 Harrls Nuclear ProJect HS Physics

),Harris Plant Engineering Section General Hanager- ED J ~ Wagner BS Hechanlcal Eng, 33 Harrls Plant Engineering Section Hanager - Harris Plant Engineering Hanagement Section H, F ~ Thompson, Jr, 'SHS Nuclear Eng.

Nuclear Eng.

Hanager - Harris Plant Engineering LE I, Loflin BS Electrical Eng, 22 BS Nuclear Eng,

TABLE 13.'I.I-I (Cont'd)

APPL I CABLE EXPERIENCE TITLE EDUCAT ION (YEARS)

SECTION OR ORGANI ZAT I ON W~eet General. Hanager- Nuclear Maritime Eng. 19 4, Hl lestone Completion Section G. Mayer BS Ml I estone Comp let ion erMe ehanie O

~ 4 Completion Assurance Section Project General Hanager - Completloq R. M. Parsons BS Civil Eng. 21 Assurance M

g. Barr.ls Plant Operations Section General Manager - Harris Plant J Wi I I ls BS Electrical Eng. 34 Operations Section

TABLE 13 ~ I 1-1 (Cont'd)

APPL I CABLE EXPERIENCE SECTION OR ORGANIZATION TITLE EDUCAT I ON (YEARS )

~.Harris ProJect Admlnlstratlon Manager - Harris ProJect M. J ~ Hlndman, Jr, BS Civil Eng. 20 Admlnlstratlon 4,Harris ProJect Planning and. Hanager - Planning and Control T. J ~ Al len BS Civil Eng, 18 Controls HBA

.Director - Planning nnd Schedul lng N L, Blair BS Engineering Mechanics 22 Manager - ProJects and Accounting H, M, Rhodes, Jr, HS Diploma 33 Q. NUCLEAR ENGINEERING AND Vice President - Nuclear Engineering A, B, Cutter BS Chem. Eng. 29 L I GENS ING DEPARTHENT and Licensing Department HS Nuclear Science 8 Eng,

1. Nuclear Licensing Section Hanager - Nuclear Licensing Section S. R Zlmmerman BS Eng, 22

3. Muclcae Ptas'plwccrig Scdiw. 4vzyr- Qw>~~PQ~f Eqineert'g Eel t(7~++Qf as re M~;~(

D.Engineering Support Nuclear Hanager - ESNPS I R. L, Sanders BS Eng. 27 PlantsSection I HS Nuclear Eng.

Q. Nuclear Engineering ProJects Hanager - Nuclear Engineering M. G. Zaalouk BS Electrical Eng. 28 Section ProJects MS Nuclear Eng.

PhD Nuclenr Eng.

+~~el-Engine

-E4eekr-4a4-

TABLE 13.1.I-I (Canted)

APPL I CABLE EXPER I ENCE SECTION OR ORGAN I ZATI ON TITLE EDUCAT I OH (YEARS )

Q. Engineering Support Nuclear Manager - ESNPS ll - J~ Hevill BS Civil Eng I8 PlantsSection I I C. Nuclear Engineering Safety Director - Huc I ear Engineering ST McManus BS Industrial Eng, Review Safety Review BS Nuclear Eng, 8, Eng, Mathematics Q.Nuclear Plant Construction Manager - Nuclear Plant Construction S~ N, Hamilton BS Science 33 Section Section I

Q. Nuclear Staff Support Section Manager - Huclear Staff Support M, D, Nl I I BS Mechanical Eng, 16 Section

- 8S F leo<rica[ E~.

~. OPERAT IOHS SUPPORT GROUP Senior VIce President Operations Support Group Nuclear Fuel Section Manager - Huclear Fuel Section L, II Martin BS Nuclear Eng 20 MBA Znzert F no~Co p ~~ IZ.].I-I~

0 ~

0

TABLE 13.1 ~ 1-1 (Cont'd)

APPL I CABLE EXPER I ENCE SECTION OR ORGAN I ZATI ON TITLE EDUCATION (YEARS)

P MATERIALS HANAGEMENT DEPARTMENT Manager - Haterlals Hanagement R, B, Richey BS Engineering 21 Department HS Industrial Eng, BhlD Y ECffhlXC l9L 4KRVXCES 8'. OPERATIONS TRAINING DEPARTMENT Vlcc President - Operations Training B, J ~ Furr BS Mechanical Eng, 23 Department I- &eery F fi cr ac'8 lh.l. I"I' I.Emergency Preparedness Section Hanager - Emergency Preparedness R, B, Black, Jr, ~

BS Industrial Eng, 15 Section Q,Nuclear Training Section Manager - Nuclear Training Section A. C. Tollison, Jr. BS Chemical Eng. 24

(:, NUCLEAR SAFETY AND ENVIRONMENTAl. Manager - Nuclear Safety and R. B. Starkey, Jr. BS Physics 21 SERVICES DEPARTMENT Environmental Services Department J,Corporate Nuclear Safety Section Hanager - Corporate Nuclear Safety J ~ D ED Jeffrles BS Eng 21 Section HS Nuclear Eng, PhD Nuclear Eng, Director - Onsite Nuclear Safety, H. W Bowles BS Physics( Eng. 14 SHNPP Director - Nuclear Safety Review J. G. Hammond BS Mechanical Eng, 18 MS Industrial Hgt.

3. Corporate Kealth Physics Section Manager - Corporate Health Physics R, L. Hayton, Jr, BS Nuclear Eng. 23 Sect ton, MS Nuclear Eng.

p. Radiological and Chemical Support Hanager Radiological and Chemical B H Webster BS Physics 27 Section Support Section

TABLE I 3. I ~ I -1 (Cont d )

~

APPLICABLE EXPERIENCE SECTION OR ORGANIZATION TITLE NAHE EDUCAT ION (YEARS )

~ ~ rt+le

..&. I<. ld~r<zer !3S ~cmm$ r~

Z. Environmental Services Section Hanager - Environmental Services Ti +i C~

I CORPORATE PUAL I TY ASSURANCE Hanager - Corporate puallty M, R. Banks MS Diploma DEPARTHENT Assurance Department Mane (

A. OA/PC Harris Plant Section cQuals'anager N. J. Chlangl MS Diploma 34 Director - PA/QC Harris Plant G. LE Forehand MS Diploma

TABLE I3.l.l-l (Cont'd)

APPL I CABLE EXPER I ENCE SECTION OR ORGANIZATION TITLE EDUCATION (YEARS )

l3. QA/QC Operations Section Manager - Operations QA/QC C, M Hoseley, Jr. BS General Eng 20 MS Nuclear/Civil Eng, Q. Quality Assurance Services Manager - Quality Assurance Services R, ED Lumsden BS Harlne Eng 32 Section Section

SllNPP FSAR TABLE 13 '.1-2 (CONT'D)

E. E. Utley Page 2 L. June 1, 1979 Appointed in charge of the Power Supply & Customer Services Groups CP&L M. May 1, 1980 Appointed in charge of the Power Supply and Engineering & Construction Groups - CP&L N. Senior Executive Vice President- Po~r 5~pp and Cowtrvc< ilaw 6 2-ough lg~~ G~gn2-cr i'g III. Professional Societies.'.

American Society of Mechanical Engineers B, North Carolina Society of Engineers C. Raleigh Engineers Club D. American Nuclear Society (National)

E. Eastern Carolinas Section of American Nuclear Society F. 'Association of Edison Illuminating Companies " Committee on Power Generation 13.1.1-19 Amendment No. ~ 3V

SHNPP FSAR TABLE 13.1 ~ 1-3 (CONT'D) fg~p. r - QUc,tea.r I Jaw+ p-rLqinWf'ipQ E. J. Wagner, .

I. Education:

A. B. S. in Mechanical Engineering from .Carnegie-Mellon University-1953 B. Additional Courses Case Western Reserve, George Washington University, and Ohio State University A. Exxon

1. Summer 1952 - Maintenance Engineer B. Babcock and Wilcox Company

l. 1953 to 1955 Test Engineer C. Naval Nuclear Propulsion Program

l. 1955 to 1970 - Chief, Nuclear Components Br'anch D. Westinghouse Electric Corporation

1. 1970 to 1975 - Division Engineering Manager E. Burns and Roe, Inc.

l. 1975 to 1983

a. Assistant to the Executive Vice President

b. Director of Engineering and Design

c. Deputy Director for Engineering d Deputy Director for Technical Evaluation F. Cincinnati Gas & Electric Company

1. 1983 to 1984 - Assistant Vice President, Nuclear Engineering G. Carolina Power & Light Company 0

1. May 1984 to January 1985 - Assistant to Executive Vice President PSE&C

2. January 1985 to Present General Manager Engineering, Harris Nuclear Project Department, Nuclear Generation Group ll~C- A~~< d~'W:o t~ zp ~,'~);e <,5

~

O~n~g r - PJuclaa~ I la~a E~ imari'<ccgg'on.. A/uc.teqy r ngineeeiq l i c.~naiad gaped.~~rig g Q'7 13.1.1-25 Amendment No.

SHNPP FSAR DELETE TABLE 13.1.1-3 (CONT'D) ines William Rhodes, Jr., Manager Project Costs & Accounting I. Education:

A. Diploma - Orrum High School, Orrum, North Carolina - 1 3 B. tended Evening Classes in Economics at North Car ina State Un ersity A. Ebasco Se vices, Inc.

1. 1951 - 1 8

a. Served i Capacities of Chic Timekeeper, Assistant Accountant and Cost Engine on Construction of Weatherspoon Unit 3, Sut n Units 1 a 2, and Cape Fear Units 5 and 6 B. Bechtel Corporation

1. 1958 1960

a. Cost Engineer Cons ruction of Port Everglades Units 1 and 2 C. Ebasco Services, nc.

1. 1960 1 0

a. Co t Engineer on Construction of ee Unit 3, Asheville it 1, Roxboro Unit 1 and Robinso Unit 2 D. Caro ina Power & Light Company 1970 to Present

a. Employed as Cost Control Specialist in the nstruction Section, Power Plant Design & Construction De rtment, located in the General Office

b. February 15, 1973 Promoted to Senior Cost Contro Specialist in the Construction Section of the Power lant Engineering and Construction Department, located in t General Office 13.1.1<<33 Amendment No. W 3P'

SHNPP FSAR TABLE 13.1.1-3 (CONT'D)

Hai s William Rhodes, Jr.

Page

c. July 21, 1973 - Promoted to Principa'1 Cost Control Specialist in the Construction Section, Power Plan Engineering and Construction Department, located 'n the General Office

d. eptember 1, 1973 - Transferred as Principal Cost Control ecialist to the Power Plant Construction epartment, lo ted in the General-Office

e. Febru ry 11, 1974 - Assigned as Princi al Cost Control Specia st to Construction Engiheeri g and Estimating Section the Power Plant Constru'on Department, located in the Ge ral Office

f. February 28, 976 Transferre as Principal Cost Control Speci'alist to onstruction S vices Section of the Power Plant Constructs n Departm t, located in the General Office

g. January 1, 1977 - ans rred as Principal Cost Control Specialist to Cost rol/Cost Reporting Section of the Engineering 6 Constr ion Support Services Department, located'n the Gen al ffice

h. January 15, 1977 - Promote to Manager, Cost Control and Cost Reporting in the Adminz trative Section of the Engineering d Construction upport Services Department, located in e General Office

i. January 0, 1981 - Transferred as irector - Cost Control 6 Cost R porting to the Power Plant C nstruction Department in the rris Site Management Section, cated in New Hill, Nor h Carolina

j. anuary 31, 1981 - Reorganization - Depar ent renamed Nuclear Plant Construction

k. December 25, 1982 - Reclassified to Manager roject Costs 6 Accounting in the Harris Site Managemen Section of the Nuclear Plant Construction Department, locat at the.

Harris Site, New Hill, North Carolina III. Professional Societies:

Raleigh Engineers Club 13.1.1-34 Amendment No. W Z7

SHNPP FSAR TABLE 13.1.1-4 Resumes of Management Personnel in the Operations Support Group Lq~U Err~

Operations Senior Vice President Support Group Education and Trainin Ete~eri ~

~

I B. S. Degree in 'ngineering, -

North Carolina State University, Raleigh, North Carolina Iq4$

A. Companies (other than CP&L) and Military Experience

. g p<; f. F740 .OMo'kc',t l.q4 0 ~.'0 5:.Prm~.

B~ 'Carolina Power & Light Company J. Summers of 1957 and'958 - Employed at Carolina Power & Light Company as a Student Summer Porker in the Northern Division Relay Office and Substation Design Office, respectively I

2. June 1959 to April 1960 - Employed as a Junior Engineez.fn the Northern ,

Division Relay Office in Raleigh, North Cazolina .

3. October 1960 through March 1970 - Employed at CP&L as a Junior Engineer in the Northern Divfsfon Relay Office in Raleigh, North Carolina, and progressed to Senior Engineer in the System Opezations Section of the Power Sup'ply Department in Raleigh, North Carolina

f. Aprfl 1970 ~"iihployed as System Operating Engineer.fn the, System Operations Section of the Generation & System Operations Department located fn the General'ffice in Raleigh,,North Carolina January 1972 through December 1976 - Employed as Manager " System Operations Section and Manager - System Operations & Maintenance Section in the Generation &. System Operations and Power Supply Departments located in the General Office in Raleigh, North Cazolina

4. January 1977 " Employed as Manager - System Operations & Maintenance in the System Operations & Maintenance Department located in the General Office in, Raleigh, North Carolina I'

June 1979 - Employed as Vice President " System Planning & Coordination Department in the Corporate Services Group located in the General Office in Raleigh, North Cazolina May 1980 thzough July 1983 - Employed as Vice President/Sr. Vice President Power Supply Group located in the General Office in Raleigh, North Carolina

g. August 1983 - Title changed to Senior Vice President " Possil Generation &,

Power Transmission. Located in the General Office in Raleigh, North Carolina lo.June 1986 - Employed as Senior Vice President - Operations Support. Located in the General Office in Raleigh, North Carolina 13.1.1-53 Amendment No. gp

SIMP FSAR TABLE 13.1.1-4 (CONT'D)

Qn~ U. Evr~

Page 2 III. Professional Societies:

A. Registered Professional Engineer -'orth Carolina & South Carolina B. ~

Institute of Electrical and Electronics Engineers C. Professional Engineers of North Carolina D. North Carolina Society of Engineers, E. American Nuclear Society F. ANS - Eastern Carolinas Section G. N. C. Chapter of the Health Physics Society

\

H. N. C. MATHCOUNTS Director I. Director - N. C. Engineering Foundati.on, Inc.

13.1.1-54 Amendment No. ~

SHNPP FSAR TABLE 13.1.1-4 (CONT'D)

Walter . Hurford, Manager Fuels Department I. Educ tion and Trainin A. B. Degree in Me'tallurgical Engineering - Carn gie Institute of Techno ogy, Pittsburgh, Pennsylvania 1942 B. S. M. Degr e in Industrial Management " Ma sachusetts Institute of Technology " Boston, Massachusetts - 196 A. 1949 1976 Manag - Light Wa r Breeder Reactor Core Activity Westinghou Bettis aboratory (Westinghouse Electric Corporation)

B. 1976 - 1981 - Vice Presi n Corporate Production - Wyoming Mineral Corporation (W tingh se Electric Corporation)

C. 1981 - 1982 Mana r of Producti << Western Zirconium Division located' (Westinghouse El tric Corporation)

' D.

January 1983 Manager -

'hechni.cal Employed at Carolina Services Department General Office, Raleigh, Pow i

Nor 6 Light I~

Company as the Power Supply Group Carolina E. Sept er 1983 - Manager " Fuels Department, Ge ral Office, Ra igh, North Carolina Pro essional Societies.'.

American Society for Metals 13.1 '-55 Amendment No. ~ZP

SHNPP FSAR TABLE 13 '.1-4 (CONT'D)

L. H. Martin, Manager ,Nuclear Fuel Section I. Education.'.

B. S. Degree in Nuclear Engineering North Carolina State University 1965 B. M. B. A. Degree University of South Carolina - 1971 A. 1965 to 1970 - Engineer in Reactor Technology Section Savannah River Plant

'B. April 1972 - July 1973 - Senior Engineer - Carolina Power 6 Light Company, Bulk Power Supply Department, Fuel Section, General Office, Raleigh, North Carolina C. July 1973 - August 1974 - Principal Engineer - Surveillance &

Accountability (In-Training) Bulk Power Supply Department, Fuel Section " CPhL

.'. August 1974 -

Accountability Bulk January 1977 " Principal Engineer - Surveillance Power Supply Department, Fuel Section -

6 CPSL

-E. January 1977 May 1977 - Principal Fuel Analyst, Fuel Department, Fuel Analysis Unit - CPGL F. May 1977 - Prese'nt - Manager - Nuclear Fuel, Nuclear Fuel Section - CPGL III. Professional Societies'.

A. Registered Professional Engineer - North Carolina 1975 B. Member of American Nuclear Society 13.1.1-56 Amendment No. ~ Z7

SHNPP FSAR TABLE 13 ~ 1 ~ 1-4 (CONT'D) czytcj 7ec4iccl( &mica'.

J. Furr, Vice President Operations Trainj.ng Department Education and Trainin A. B. S. Degree in Mechanical Engineering North Carolina State University " 1962 B. Basic Surveying Course - 1965 C. Basic Radiological'ealth Course Conducted by the Public Health Service, Winchester, Massachusetts '- 1966 D. Reactor Safety and Hazards Evaluation Conducted by the U. S.

Public Health Service, Rockville, Maryland 1968 E. Westinghouse Nuclear Reactor Training Program 1968 F. Senior Reactor Operator License on H. B. Robinson A. June 1955 to July 1958 U. S. Army - Instructor in Aviation Maintenance Summer 1960 - Summer Student Worker Substation Shops Carolina Power & Light Company - Raleigh, North Carolina C ~ Summer 1961 - Summer Student Worker - Cape Fear Steam Electric Plant Carolina Power & Light Company Moncure, North Carolina D. June 1962 to May 1963 Engineer E. I'uPont de Nemours Company E. May 1963 Employed as a Junior Engineer at the W. H. Weatherspoon Plant, Lumberton, North Carolina F. February 1964 Employed as a Junior Engineer at the H. B.

Robinson Plant, Hartsville, South Carolina G. July 1964 Employed as a Mechanical Engineer at the H. B.

Robinson Plant, 'Hartsville, South Carolina H. January 1966 Employed as a Mechanical Engineer at the Roxboro Steam Electric Plant, Roxboro, North Carolina I. February 1966 " Employed as Operating & Results Supervisor at the H. B. Robinson Plant, Hartsville, South Carolina 13 ~ 1 ~ 1-59 Amendment No. A; 8'7

SlBPP FSAR TABLE 13.F 1-4 (CONT'D)

BE J. Furr gage 2 J. Septembez 1971 - Employed as a Principal Engineer in the Nuclear Generation Section of the Generation & System Operations Department in the General Office K. June 1972 - Employed as Plant Superintendent 'in the Nuclear Generation Section of the Generation 6 System Operations at the H. B. Robinson Plant, Hartsville, South Carolina

'epartment L. July 1974 - Employed as Manager - Nuclear Generation Services in the Nuclear Generation Section of the Bulk Power Supply Department in the General Office M. May 1976 Employed as Plant Manager II (Temporary) in the Nuclear Generation Section of the Bulk Power Supply Department at Brunswick Steam Electzic Plant, Southport, North Carolina N. December 1976 Employed as Manager Nuclear Generation Services in the Nuclear Generation Section of the Bulk Power Supply Department in the General Office

0. January 1977 - Employed as Manager Generation Department. in the Power Supply Group in the General Office

-P. October 1979 - Employed as Manager Nuclear Operations in the Power Supply Group in the General Office

'. December 1979 - Employed as Vice President - Nuclear Opezations the Power Supply Group in the General Office in R. September 1983 - Employed as Vice President - Operations Training 6 Technical Services Department in the Operations Support Gzoup in the General Office S. October 1985 - Employed as Vice President Operations Training Department in the Operations Support Group in the General Office Rages~ 1<~< Em ptoggmd os< Vice. F}a@i Jand -Openzfgons T~t~ig.

T~&nicsc(Senr~ ceg Pepvfr'can+.

~~

professional Societies:

A. Member of American Society of Mechanical Engineers B. Member of American Nuclear Society 13.1.1-60 Amendment No. ~ g7

SHNPP FSAR TABLE 13.1.1-4 (CONT'D)

A. C. Tollison, Jr.

Page 2

7. 1981 to 1983 On loan to Institute of Nuclear Power Operations

a. 1981 Evaluator, Evaluation Team Manager Manager Organization & Administration Department

b. 1982 to 1983 - Director Evaluation & Assistance Division

8. September 1983 - Employed as Manager << Nuclear Training, Operations Training & Technical Services Department, Shearon Harris Energy & Environmental Center, New Hill, North Carolina

9. October 1985 Employed as Manager - Nuclear Training, Shearon Harris Energy &

Environmental Center, New Hill, North Carolina III. Professional Societies.'.

American Nuclear Society

13. l. 1-62 Amendment No. & g7

SHNPP FSAR TABLE 13.1.1-4 (CONT'D)

Robert G. Black, Jr., Manager Emergency Preparedness Section Education and Trainin A. B. S. Degree in Industrial Engineering - Georgia Institute of Technology 1969 B. Attended various schools while i'n the U. S. Navy C. Registered Professional Engineer - February 1979

~ . A. June 1969 to June 1973 - V. S. Navy Nuclear Program B~ September'973 - Senior Engineer - Environmental & Technical Services Section Special Services Department, CP&L, Raleigh, North Carolina January 1976 to June 1976 - Project Engineer >> Licensing &

r C~

Technological Services Section, Special Services Department, CP&L, Raleigh, North Carolina L D, June 1976 to December 1979 - Project Engineer - Nuclear Licensing Unit, Licensing & Siting Section, Technical Services Department, CP&L, Raleigh, North Carolina E. December 1979 to Ma'rch 1981 - Project Engineer - Nuclear Licensing Unit, Licensing & Permits Section, T'echnical Services Department, CP&L, Raleigh, North Carolina F. March 1981 to August 1983 " Director - Emergency Preparedness-Technical Services Department, General Office, Raleigh, North Carolina G ~ August 1983 to May 1985 " Director Emergency Preparedness Unit, Operations Training & Technical Services Department, General Office, Raleigh, North Carolina May 1985 to October 1985 - Manager Emergency Preparedness Section, Operations Training & Technical Services Department, General Office, Raleigh, North Carolina I ~ October 1985 to Present Manager Emergency Preparedness Section, General Office, Raleigh, North Carolina 13.1.1"63 Amendment No. W ZF

SIQG'P FSAR TABLE 13.1.1-4 (CONT'D)

R. B. Starkey, Jr.

Page 2 July 1972 - September 1973 Lt. Commander, U. S, Navy Assistant Director of the Engineering Division - U. S. Navy Submarine School F. September 1973 " February 1974 Senior Engineer Nuclear Generation Section, Bulk Power Supply Department, Carolina Power &

Light Company, Raleigh, North Carolina G. February 1974 - February 1975 Senior Engineer, Quality Assurance Section, Bulk Power Supply Department, Carolina Power & Light Company, Raleigh, North Carolina H. February 1975 - Apx'il 1975 " Principal Engineex', Quality Assurance Section, Bulk Power Supply Department, Carolina Power & Light Company, Raleigh, North Carolina April 1975 - May 1976 - Quality Assurance Supervisor, Nuclear Generation Section, Bulk Power Supply Department, Carolina Power &

Light Company, Brunswick Plant, Southport, North Carolina May 1976 - December 1976 Superintendent - Technical and Administrative, Nuclear Generation Section, Bulk Power Supply.

Department, Carolina Power & Light Company, Brunswick North Carolina Plant,'outhpox't, K. December 1976 - November 1977,- Superintendent " Opexation and Maintenance, Nuclear Generation Section, Generation Department, Carolina Power & Light Company, Brunswick Plant, Southport, North Carolina L. November 1977 - Nov'ember 1979 - Plant Manager H. B. Robinson Plant, Nuclear Generation Section, Generation Department, Hartsville, South Carolina (CP&L)

Mo November 1979 " September 1983 - General Managex - Robinson, Nuclear Operations Depaitment, Carolina Power & Light Company, Hartsville, South Carolina N. September 1983 - April 19 Nuclear Operations at .th ga

'left ble CP&L) Executive Directox Hill Nuclear Plant (under construction), Public Se ' indiana 0~ April 1984 October 1985 Manager Environmental Services, Environmental Services Section, Operations Support Group, Carolina Power & Light Company, Harris E&E Center, New Hill, North Carolina Po October 1985 Present - Manager Nuclear Safety & Environmental Services Department, Operations Support Group, Carolina Power &

Light Company, Raleigh, North Carolina

13. 1. 1-66 Amendment No. X 5'7

SHNPP FSAR TABLE 13.1.1-4 (CONT'D)

J. D. E. Jeffries Page 2 J. April 1978 - Employed as Principal Engineer, Nuclear Safety, CNShQAA Section, System Planning 6 Coordination Department, Carolina Power & Light Company, Raleigh, North Carolina K. June 1979 - August 1981 - On loan to the Electric Power Research Institute as Project Manager, Nuclear Division, Palo Alto, California L. August 1981 - Employed as Manager - Corporate Nuclear Safety Section, Carolina Power 6 Light Company, Raleigh, North Carolina III. Professional Societies'.

A. American Nuclear Society B. Society of the Sigma Xi C. Health Physics Society North Carolina Section D. Registered Professional Engineer in North Carolina and Pennsylvania 13.1.1-69 Amendment No. Q Z'P

SHNPP FSAR

! TABLE 13.1.1-4 (CONT'D)

H. W. Bowles Page 2 H. September 1981 November 1982 - Employed as Project Engineer, Corporate Nuclear Safety Section, Corporate Nuclear Safety &

Research Department, Carolina Power & Light Company, Raleigh, North Carolina I.

Safety (SHNPP), Corporate Nuclear Safety Section, Company, Raleigh, North Carolina

~~

November 1982 Present - Employed as Director - Onsite Nuclear Carolina Power & Light III. Professional Societies:

A. American Nuclear Society B. Professional. Engineer of North Carolina

13. 1. 1-71 Amendment No. ~07

SHHPP FSAR TABLE 1301 '-4 (CONT'D)

B. H. Webster Page 2 L. February 1982 Employed as Manager - Radiological & Chemical Support Section in the Technical Services Department. (Located at the Harris Energy & Environmental Center) - CP&L January 1981 Employed as Manager - Environmental & Radiation M.

Control in the Technical Services Department. (Located at the Harris Energy & Environmental Center) - CP&L III. Professional Societies'.

A. North Carolina Chapter - Health Physics. Society B. American nuclear Society of East Carolina Section C. Power Reactor Health Physics Group 13.1.1-75 Amendment No. + g/

SHNPP FSAR TABLE 13.1.1-5 Resume of Vice President Brunswick Nuclear Project PE W. H e, Vice President - Brunswick Nuclear Project I. Educ tion.'.

Bac elor of Science Degree in Chemistry from The itadel,,

Char eston, South Carolina in 1951 B.'ertifi ate Engineering Management. UCLA 1963 C. Member of .S.A.E.C., Atomic Safety & Lice sing Board from 1962-1966 A. September 1951 to ebruary 1956 aboratory Supervisor <<

E. I. du Pont de Ne ours & Compa y, Inc., Savannah River Plant, Aiken, South Carolin B. February 1956 to August 195 Senior Nuclear Engineer The Martin Company, Nuclear ision, Baltimore, Maryland C. August 1956 to August 1 7 Superintendent - Olin Mathieson Chemical Company, Nuc ar Fue s Division, New Haven, Connecticut D. August 1957 to Jun 1966 Depar ment Head Lawrence Radiation Laboratory, Unive sity of Califor a, Berkely, California E. September 1967 o March 1971 " Chief, Site Environmental and Radiation Sa ty Group - Division of actor Licensing, U. S. Atomi Energy Commission, Washing on, DC F. March 19 to November 1971 - Manager En ironmental & Technical Servic Section of the Generation & System perations Department, Carol a Power & Light Company G. No ember 1971 to February 1974 Manager- Enviro mental &

chnical Services Section, Special Services Depa ment - CP&L H. February 1974 to February 1975 Manager - Licensing Technological Services Section, Special Services Depa ment CP&L I. February 1975 Manager Special Services Department, Engineering, Construction & Operation Group CP&L J. June 1976 Manager - Technical Services Department, Engineer ng, Construction & Operation Group - CP&L 13.1.1-76 Amendment No.

SIBPP FSAR TABLE 13.1 ~ 1-5 (CONT'D)

P. W. Howe Page 2 0

K. December 1976 President T nical Services Department, r l Engineering & Construe n Gr - CP&L L. October 1982 - Vice esident - unswick Nuclear Project - CP&L pt(

III. Professional S eties:

A. rican Nuclear Society

13. 1. 1-77 Amendment Nn.'~ +~7

SHNPP FSAR TABLE 13.1.1-6 (CONT'D)

I H. R. Banks Page 2 I June 1960 -. June 1962 Chief Engineman, EOOW, Nuclear Power Training Unit, (SIW) Nuclear Submarine Prototype - Idaho Falls,

~

Idaho J~ June 1962'- October 1964 USS Andrew Jackson, SSBN 619 EOOW, Leading Machinery Division Chief, Supervisor in charge of operation of the nuclear power plant K. October 1964 January 1965 --Naval Officer's Candidate School L. January 1965 August 1968 - Nuclear Ship Superintendent San Francisco Bay Naval Shipyard M. August 1968 July 1970 - Resident Project Engineer " H. B.

Robinson Plant - Unit No. 2 CP&L, Power Supply Department, Hartsville, South Carolina N. July 1970 - August 1971 " Resident Project Engineer Brunswick Plant - Units 1 & 2 - CP&L, Power Plant Design & Construction, Department, Southport, North Carolina

0. August 1971 " February 1972 " Manager Quality Assurance, Power Plant Design & Construction Department, CP&L, Raleigh, North Carolina P. February 1972 - July 1973 Manager Quality Assurance Audit, Special Services Department - CP&L, Raleigh, North Carolina July 1973 August 1975 - Manager Quality Assurance & Training Q~

Audit, Special Services Department - CP&L, Raleigh, North Carolina R. August 1975 -. March 1976 Manager Corporate Quality Assurance Audit, Special Services Department - CP&L, Raleigh, North Carolina So March .1976 " October 1979 - Manager Nuclear Generation, Bulk Power Supply Department, Nuclear Generation Section - CP&L, Raleigh, North Carolina,

- Fe.iraq /'7~I To October 1979)- Genera Manager Shearon Harris Nuclear Power Plant - CP&L, Raleigh, North Carolina U. February 1981 - Present - Manager Corporate Quality Assurance-CP&L, Raleigh, North Carolina III. Professional Societies:

A. ASME Standards Committee Main Committee NQA, Subcommittee Operations N45-2.12 & N45.2.23 B, EEI QA Committee Ce American Society of Nondestructive Testing Do American Society of Mechanical Engineers E. American Nuclear Society F~ North Carolina Society of Engineers.

13.1.1-79 Amendment No. A. >7

SHNPP FSAR TABLE 13 '.1-6 (CONT'D)

W~aar<a.( QM)'k~

N. J. Chiangi, Manager I ~ Education:

A. Graduate of Norwich Fxee Academy, Norwich, Connecticut Bo Special Schools: Nuclear Submarine Systems, Navyships 250-1500-1, Mil. Std. 271 D-271A, Navyships 250-693-1 693-3 (structural),

Health Physics Monitoring, Management Scho'ols - Electric Boat Company, Electronics School "-U. S. Navy, Welding School - EBC, Radiography School, Magnetic Particle Testing School EBC, Liquid Penetrant Test School - EBC, Ultra Sonic Testing Classes -, EBC, Eastman Kodak School for Automatic Film P'rocessing Equipment, Job Cost Estimating EBC.'ualified: AEC Licensed Radiographer and Radiographer Supervisox A. 1947 - 1952 U. S. Navy, Sonar Man - Radar Man. Special Training, Electronics School, Sonar School, Radio School B~ 1952 - 1967 Electric Boat Company, Groton, Connecticut

' l.

2.

1952 - 1954 1954 1967

"'elding-Field Work-Piping-Stxuctural Lead-Supervisor - Radiography Department.

Responsible for all Nuclear Radiography Structural-Piping-Castings, Polaris Missile Program, Radiographer, Film Readers. Setup, wrote, and reviewed Radiography, Test Procedures for Casting-Piping-Structural Radiography.

Instructed Piping and Mechanical design personnel, instructed Radiogxaphy Classes for New Hires, reviewed an3 interviewed personnel for hire. Attended Management-Quality Contral meetings.

C. 1967 - 1973 - Ebasco Services, Inc., New York, New York

l. 1967 - 1970 - Quality Compliance-Quality Control Supervisor for Ebasco at H. B. Robinson NPS Unit No. 1. Responsible fox implementation of the site for the H. B. Robi'nson pr'oject.

This included supervising Ebasco site Quality Compliance Representatives in the performance of their inspection duties in the following areas: welding, civil, electrical, nondestructive testing, receiving, storage, and testing.

Responsible for the review of'site purchase orders for quality requirements and documentation to assuxe its adequacy.

Responsible for maintaining Quality Assurance documentation.

13.1.1"80 Amendment No ~ W >7

SIMP FSAR TABLE 13 ~ 1 ~ 1-6 (CONT'D)

N. J. Chiangi Page 2

2. 1970 1972 Site Quality Compliance S'upervisor for Ebasco at St. Lucie No. 1 Nuclear Power Plant, with responsibility for implementing the site phase of the Ebasco Quality Program as modified for St. Lucie. Responsible for auditing field construction activities as required by the Quality Program, auditing the performance of construction quality control tasks through'he Site Quality Compliance Staff, meeting with AEC representatives in performance of their site audits, and maintaining quality compliance files as described in the Ebasco Quality Program for representation to the client at the completion of the pr'oject.

3. 1972 - 1973 - Senior Quality Compliance Engineer for Ebasco at Chin-Shan Unit Nos ~ 1 and 2. Had overall responsibility for Ebasco Quality Compliance Program on site. Duties at Chin-Shan site included the following.'nstructed personnel in inspection of welding, mechanic@1, civil, and electrical functions. Responsible for interpretation of all codes and specifications having to do with this project where compliance or. contr'ol was required. Instructed and trained Ta'ipower Personnel in Quality Compliance and Quality Control functions. Developed quality control and compliance programs for Taipower. Responsible for a vendor inspection.

Interpreted all radiographs on site. Responsible for maintaining radiographs and quality assurance documentation.

D. October 1, 1973 - Carolina Power 6 Light Company, Raleigh, North Carolina - Employed as Quality Assurance Manager - Construction, Quality Assurance Section of the Power Plant Construction Department, located in the General Office E. November, 1976 - Manager, Engineering and Construction QA Section, Technical Services Department - CP&L F. March 1983 - Manager, QA/QC Harris Plant Section of the Corporate Quality Assurance Department - Harris Site, New Hill, North

. ~ Carolina fail(

III. Professional Societies!

A. Member - ASNT - ASME B. Qualified ANST Level III - 2/4/77 Radiographic - Magnetic Particle - Liquid Penetrate C. Professional Engineer - State of California - January 1977 ro.. Wurra I iaaf fgrrrarar -iSa'aria.( rsiuai kgZac+r'ov 'ef klrrrrga tarq+

QucJ ig RSsur'rrrrae13-~ roan+ .Naru NorVLCarolirra 13.1.1-81 Amendment No. A 37

SIiNPP FSAR TABLE 13.1.1-6 (CONT'D)

Cha les Lewis McKenzie, Jr., Principal Quality- Assurance Engineer - ~

r Acting Director QA/QC Harris (Operat's) ucation and Trainin A. . S. Industrial Engineering, University of Flor' - 1971 B. Su lemental at Charleston Naval Shipyard

1. N lear Cleanliness', Reactor Fundament s, 11/71'- 1/72

2. Nucl r Quality Control - 9/72 - 12/ 2

3. S5W Rea tor Plant - 1/73 5/73 C; Basic Ultrason c Testing Course pr sented by Magnaf lux Corporation - 1 5 9/73 D~ Orientation of New Appointe Supervisors presented by CPGL, General Office 11 8 20/

E~ . Basic Principles of Sup r xsory Management presented by CPhL.

General Office 6/8 76 erience'.

A. U. S. Post Office

1. April 1966 o December 1967

a. Pos al Clerk B. Pratt 6 W itney Aircraft-Turbine Engine Division

1. Ju e 1968 to March 1969

a. Engineering Aid Co-Op Student C. ates 6 Daily Construction Company

1. June 1969 to September 1970

a. Laborer (summer employment while attending c lege) 13.1 ~ 1-86 Amendment No. Z7

SHNPP FSAR TABLE 13.1.1-6 (CONT'D)

Cha es Lewis Mckenzie, Jr.

Page D. Charleston Naval Shipyard August 1971 to August 1973 Nuclear Auditor for Naval S5W Overhaul and efueling Pro g ram E. Carolina Power Ec Light Company Septem r 1973 - Employed as a Junior Qu ity Surveillance Speciali t in the Quality Assurance Sec ion of the Power Plant Engineeri Department, located in th General Office, Raleigh, No th Carolina

a. October 1 74 - Promoted to V ndor Surveillance Specialist in the Qua 'ty Assurance S tion of the Power Plant Engineering . epartment, 1 cated in the General Office, Raleigh, Nort Carolina

b. April 1976 - Pro ote to Vendor Surveillance Specialist I'II in e Quality Assurance Section of the Power Plant Engin ing Department, located in the General Office, ale h, North Carolina C ~ November 1976 Transfe ed as a Vendor Surveillance Specialist I in the Eng neering 6 Construction Quality ection of the T chnical Services Department, Assurance

'located 'he General Offi , Raleigh, North Carolina d~ June 78 Promoted to Senior endor Surveillance Spe 'alist in the Engineering 6 onstruction Quali<y As rance Section of the Technica Services Department, 1 cated in the General Office, Rale'gh, North Carolina

e. August 1980 - Reclassified to Senior ality Assurance Engineer in the Engineering 6 Construct'on Quality Assurance Section of the Technical Servi s Department, located in the General Office, Raleigh, No th Carolina March 1981 >> Transferred as a Senior Quality ssurance Engineer in the Engineering 6 Construction Qua ity Assurance/Quality Control Section of the Corpor e Quality Assurance Department, located in the Gene al Office, Raleigh, North Carolina 13.1.1-87 Amendment Nn.

SIMP FSAR f es Lewis Mckenzie, Jr.

TABLE 13.101-6 (CONT'D)

Cha Page

g. August 1981 Promoted to Project Quality Assurance/Quality Control Engineer in the ngineering &

Construction Quality Assurance/Quality ontrol Section of the Corporate Quality Assurance Depa ment, located in he General Office, Raleigh, North arolina

h. Feb ary 1983 SECTION TITLE ANGE Project QA Engine r in the QA Engineeri g Unit of the Quality Assuranc Services Sectio .of the Corporate Quality Assurance partment, 1 ated in the General Office, Raleigh, Nor Caroli March 1983 Pro ed to Principal Quality Assurance Engineer in the A ngineering Unit of the Quality Assurance,Ser aces Se tion of the Corporate Quality Assurance partment, 1 cated in the General Office, Raleigh, orth Carolina

j. May 85 - Transferred as a P 'ncipal QA Engineer (Acting Di ector >> QA/QC Harris (Operat ns)] in the QA/QC Harris Operations) Unit of the Operatio QA/QC Section of the Corporate Quality Assurance Departme t, located in the General Office, Raleigh, North Carolin III. Prof ssional Societies:

American Society for Quality Control B. Registered Professional Engineer in the State of North Caro 'na-February 1976 13.1.1-88 Amendment No. + g/

SHNPP FSAR 13ol ~ 2 OPERATING ORGANIZATION 13.1.2.1 Introduction The SHNPP organization is based on the considerable expe'rience that CPGL has operating its three nuclear units, Robinson Unit No. 2 and Brunswick Units 1 and 2. Carolina Power & Light Company will comply with ANSI N18.7-1976, "Administrative Controls and Quality Assurance for the Operational Phase of .

Nuclear Power Plants," as indicated in Section 1.8, in the operation and administration of the Shearon Harris Nuclear Power Plant. The succession of responsibility in the event of absences, incapacitation of personnel, or other emergencies are outlined by the organization chart (Fig. 13.1.2-1), The staff loading schedule is shown in Table 13.1.2-1.

13 F 1.2.2 Personnel Functions Res onsibilities and Authorities 13 '.2.2.1 Plant General Manager - Harris Plant Operations Sacro<

The Plant General Manager is responsible for all phases of plant management, including administration, opera'tion, maintenance, and technical support. He manages .and controls the organization through personal contact with the Assistant Plant General Manager and eevea-unit heads and through written ) argy reports, meetings, conferences, and in-plant inspections. He is responsible for adherence to ~requirements of the operating license, technical (sr specifications, Corporate Quality Assurance Program, and Corporate Health Physics and Nuclear Safety policies. He is responsible for reviewing incoming and outgoing correspondence with the NRC Office of Nuclear Reactor Regulation and the Office of Inspection and Enforcement concerning the Harris Plant; the establishment and approval of qualification'equirements for all Harris Plant Operations staff positions'the personal review of the qualifications of specific personnel for managerial and supervisory positions in the. Harris Plant Operations Section; and the. review of and concurrence in the plant radiation protection, radiological security, quality assurance, fire protection, training, operations, and maintenance programs. He is supported in these responsibilities. by the Assistant Plant General Manager, Director/-

Plant Programs and Procedures, Manager - Maintenance, Manager Environmental and Radiation Control, Manager'- Operations, Manager Technical Support, the Manager - Startup and Test, and Director, " Regulatory Compliance. He has the authority to issue procedures, standing orders, and special orders. In the absence of the Plant General Manager, the Assistant Plant General Manager assumes his authority and responsibilities. The Plant General Manager reports directly to the Vice President - Harris Nuclear Project Department.,

13 F 1.2.2.2 Assistant Plant General Manager

'I The Assistant Plant General Manager has the responsibilit'y and accountability for the safe, reliable, and efficient daily operation of the Harris Plant. He has direct control over the operations, maintenance, environmental/chemistry/

radiation control, and technical support functions. He is responsible for adherence to ak1 requirements of the Operating License, Technical Specifications, Corporate Quality Assurance Program, and Corporate Health

<he Further information is contained in the TMI appendix. ~ I,

SHNFP FSAR Physics and Nuclear Safety policies. He is responsible f'r the personal review of the training and qualification requirements of the following managers who report directly to him: Manager Operations, Manager Maintenance, Manager Environmental and Radiation Control, and Manager Technical Support.

Plant General Manager'.

• "'-'"" ~ I 13.1.2.2.3 Plant Programs and Procedures Unit The Plant Programs and Procedures Unit provides support functions such as security, procedure control, and emergency preparedness.

cern [ The Director/-Plant Programs and Procedures provides direct support to the Plant General Manager in the areas of security, emergency preparedness, procedure development and control, personnel administration and plant administrative coordination', directs plant security planning and activities; directs emergency preparedness planning and activities at the plant staff level; supervises the preparation, review, approval and distribution of plant procedures and directives. He is assisted in these duties by ~cz Security Supervisor, and a Senior Specialist Emergency Preparedness. The Director, Plant Programs and Procedures reports The Sanicy 5 ec,i'ass'~4- ~

to the Plant General Manager Harris Plant.

F'la~

w &e.)or $pscMi'<4-Pla.~u~d P~g~m<,

Propre.~ prev)'et+a the administrative func'tions of the plant including incoming correspondence screening and action assignment; action item/response development and follow-up; outgoing correspondence procedure preparation, review, and approvalg The Security Supervisor develops, implements, and maintains a security program

.which ensures that the security of the plant is maintained in accordance with NRC requirements. He maintains a close working relationship with local law enforcement agencies to ensure compliance with NRC regulations. He provides input to the Training Unit so that employees requiring access to the plant are properly trained and badged. He ensures that equipment and guards are available and in a state of readiness; The Senior Specialist Security is assisted by Technical Aides and a contract security 'guard force. The Security Supervisor reports to the Directorj-Plant Programs and Procedures.

The Senior Specialist Emergency Preparedness is responsible for the continuing refinement of the plant Emergency Preparedness Program which ensures that a "state of readiness" is maintained at the plant to cope with any classification of emergency. He incorporates the provisions of the plant Emergency Plan in the program and,rdvises the program and related procedures as changes are made in the plant Emergency Plan. He coordinates the training of Technical Support Center participants and the annual Emergency Drill. The Senior Specialist Emergency Preparedness reports to the Director/-Plant Programs and Procedures.

SIHIPP FSAR 13.1.2.2.3.1 'nvironmental 6 Radiation Control Unit The Manager - Environmental 8 Radiation Control (E&RC) is responsible for the plant 'radiation safety and control (health physics) programs, the plant chemical control programs, and the environmental programs. These programs are designed to ensure that environmental and radiation control is maintained in a manner which will protect the plant, employees, visitors, general public, and the surrounding community. He has the authority to issue special orders. His primary responsibility is organizing, planning, and controlling ESRC resources to provide the required support while ensuring compliance with plant Technical Specifications, the ALARA concept, and all applicable state and federal regulations and permit requirements.

Some of his major responsibilities include: (1) ensuring that programs and related procedures are developed and administered to.meet plant needs and regulatory requirements,'2) maintaining an awareness of current and pending regulations in the areas of radiation control, chemistry, and environmental matters concerning plant operations; and (3) providing adequate documentation pertaining to individual radiation exposures, radioactive effluents, chemical control of plant systems and environmental surveillance and ensuring that these records are maintained in an up-to-date, retrievable manner. He is assisted in these functions by an Environmental 6 Chemistry Supervisor, a Radiation Control Supervisor, a Project Specialist - Environmental and Chemistry, a Project Specialist - Radiation Control, and a staff of radiation control and environmental and chemistry specialists, foremen, and FvM.

technicians.'he Manager reports to Plant General Manager. The Managery -f+Ra, the'ssistant

~h+e4, has direct access to the Plant General Manager for matters relating to radiological health and safety of employees and the public.

The Environmental 6 Chemistry Supervisor plans, organizes, and directs chemistry control and environmental surveillance programs, maintains laboratory procedures, test results and records, and adheres to 'the requirements of the operating license and technical specifications'e accomplishes these responsibilities 'through foremen and technicians'he Unit.

The Radiation Control Supervisor is responsible for the plant Radiation Control (Health Physics) Program and for ensuring that @44- plant activities are conducted in a manner which will protect the plant, employees, visitors, general public, and the surrounding community. His primary responsibility is organizing, planning, and controlling Radiation Control Subunit resources to provide the required support while ensuring compliance with plant Technical Specifications and all applicable state and federal regulations and permit requirements. He accomplishes this through foremen and radiation control technicians. The Radiation Control Supervisor reports to the Manager - E+R<

Unit.

The Project Specialist << Environmental 6 Chemistry provides technical advice and recommendations for program enhancement to the Manager EGRC, and ensures that the Environmental and Chemistry Programs support efficient, reliable 13 '.2-3 Amendment ETo. + ZP

SHNPP FSAR plant'perations. He is the Environmental Chemistry technical expert for the Manager - EERC. He is supported by a staff of specialists and technicians and reports to the Manager Unit.

E+ RC.

The Project Specialist Radiation Control provides technical advice and recommendations for program enhancement and ALARA prog'ram considerations to the Manager ESRC, and ensures that the Radiation Control Programs support efficient and reliable plant operations. He is the Radiation Control technical expert for the Manager - E&RC. He is supported by a staff of specialists, technicians, and clerks and reports to the Manager-Environmental and Radiation Control Unit.

13.1.2.2.3.2 Maintenance Unit The Maintenance Unit performs ~ corrective and'reventive maintenance on plant systems and equipment. The Manager Maintenance is responsible for corrective and preventive maintenance for the equipment of the unit and in the support facilities. This includes ensuring that the equipment and associated instrumentation and controls, mechanical, and electrical systems in 'the unit and support facilities are maintained at optimum dependability and operating efficiency. He is responsible for the coordination of these functions and for approval of Special Orders, working procedures and standards. He is assisted by the Mechanical Maintenance Supervisor, Electrical Maintenance Supervisor, Project Engineer - Maintenance, Project Specialist " Maintenance, Project Engineer Computer, and a staff of engineers and specialists, foremen, mechanics, electricians, painters/pipe coverers, planner/analysts, and technicians. The Manager - Maintenance reports to the Assistant Plant General Manager - Operations.

The Maintenance Supervisor Electrical ensures that equipment, instrumentation, controls, and electrical systems are maintained at optimum dependability, safety, and operating efficiency to comply with plant technical specifications, QA, Security, Radiation Control and plant procedures, and regulatory requirements'e accomplishes this by planning, directing, and controlling a trained staff, inspecting maintenance work, providing effective maintenance procedures and standards, and developing improvements in the Preventive and Corrective Maintenance Program. He is assisted in these functions by a staff of foremen, technicians, and electricians. The Maintenance Supervisor - Electrical reports to the Manager - Maintenance Unit.

The Maintenance Supervisor Mechanical ensures that mechanical systems are maintained at optimum dependability, safety, and operating efficiency to comply'with plant technical specifications, QA, Security, Radiation Control,

~

and plant procedures> and regulatory requirements. He is responsible for all required painting and pipe covering activities necessary to maintain neat, properly insulated plant systems'e accomplishes this by planning, directing, and controlling a trained staff, inspecting maintenance work, providing effective maintenance procedures and standards, and developing improvements in the Preventive and Corrective Maintenance Programs. He is assisted by a staff of foremen, mechanics, and painter/pipe coverers. The Maintenance Supervisor - Mechanical reports to the Manager Maintenance

&zbani-h .

th)+

13.1.2-4 Amendment No. ~ g,7

SINPP FSAR The Pxoject Engineer Maintenance provides technical support to plant electrical and mechanical maintenance and assists the Manager Maintenance in assuring that plant instrumentation, control, electrical systems and mechanical systems are maintained at optimum dependability, safety, and operating efficiency, and remaining in compliance with ash technical specifications and regulatory requirements'e is responsible for administration of the Maintenance Management System to accomplish the planning and scheduling of maintenance, ensuring parts availability, and establishing clearances necessary for preplanned work; he is assisted by a staff of engineers, specialists, technicians, and planner/analysts. The Project Engineer Maintenance reports to the Manager - Maintenance Unit.

The support and technical expertise to ensure that ~

Project Engineer Computer provides process computer system maintenance plant process computer systems are fully operational for the safe, reliable, and efficient operation of the plant. He is assisted by a staff of specialists and technicians. The Project Engineer - Computer reports to the Manager " Maintenance Unit.

13.1.2.2.3.3 Operations Unit The Manager - Operations ensures that the safe and efficient operation of the unit and required support facilities'e is responsible for primary and secondary system performance and the timely completion of the scheduled periodic tests, and for adherence to the requirements of, the operating license and technical specifications. He is also responsible for coordinating and overseeing the duties of the Operating Supervisorsassigned to the plant, the Radwaste Supervisor, and the Operations Support Supervisor. He is responsible'or orderly and safe operations, turnovers, and compliance with operating instructions. He shall hold a 'icense. He has the authority to issue Special Orders. He is supported in these responsibilities by a staff of the Operating Supervisor, Radwaste Supervisor, Operations Support Supervisor, engineers/specialists, Shift Technical Advisors, Shift Foremen, and Operators. The Manager Operations Unit reports to the Assistant Plant General Manager.

The Operating Supervisor supervises plant operations. He is responsible for adherence to the requirements of the Operating License and Technical Specifications. The Operating Supervisor is responsible for scheduling and reviewing surveillance tests, reviewing operating data, logs and records, shift reports of equipment malfunctions or unusual system behavior, and initiating corxective action. The Operating Supervisor reports to the Manager " Operations. The Operating Supervisor shall hold an SRO License.

The Harris Plant Operations Section will have shift opexating crews assigned to provide 24-hour coverage of plant activities'ach shift operating crew will be manned in accordance with Technical Specification Table 6.2"1.

Each Shift Operating Crew in the Harris Plant Section shall meet the following requirements:

a) When the unit has fuel in the reactor core, there shall be a Shift Foreman with an SRO license on site at all times.

13.1.2-5 Amendment No. Ã R7

SlQIPP FSAR b) When the unit has fuel in the core, there shall be a licensed operator in the control room at all times.

c) control When room licensed operator in the "f~

the reactor is operating, there shall also be at all times.

p C)Pc:Pcl f l~

a licensed control room to provide relief for the control room operator and to perform duties outside the control room that need to be performed by a licensed operator.

SRO in the e) When the reactor contains fuel, there shall be an auxiliary operator in addition to the individuals required in (a) through (d) above. An additional auxiliary operator is required for the-control room when the reactor is be+ay eyer ~4 ~~ca<i~.

f) For all core alterations, there shall be a licensed,SRO or SRO limited to Fuel Handling to directly supervise the core alteration. This SRO shall not be assigned any other concurrent operational duties.

g) The Shift Foreman shall be assigned only the minimal administration duties required to operate his shift.

An extensive training program has been established to ensure that each onsite crew collectively has the requisite technical qualifications in reactor physics and control, nuclear fuel, thermal hydraulics, transient analysis, instrumentation and control, mechanical and structural engineering, radiation control and health physics, electric power, chemistry, and plant ope'ration and maintenance.

The Shift Foremen ensure the safe, dependable, and efficient operation of the plant during their assigned shift. ~ ~

They are responsible for adherence to the operating procedures, the operating license, and technical specifications.

is the responsibility and authority of the Shift Foreman to maintain the broadest perspective of operational conditions affecting the safety of <he Room duty. The Shift Foreman shall hold a Senior Operator's license. The "

Shift Foreman, until properly 'relieved, remains in the Control Room at all times during an accident to di:rect the activities of Control Room Operators.

He may be relieved only by qualified persons holding SRO licenses'uring routine operations when the Shift Foreman is temporarily absent from the Control Room, a Senior Control Operator will be designated to assume the Control Room command function. He is supported by and supervises Senior Control Operators, Control Operators, and Auxiliary Operators. The Shift Foreman reports to the Operating Supervisor.

The Shift Foreman is the designated individual in charge of the plant on back gl shifts unless specifically relieved of the responsibility by either the Operating Supervisor, Manager Operations, Assistant Plant General Manager, or the Plant General Manager. They are responsible for all personnel assigned on the back shifts including operators, mechanics, electricians, RC technicians, and IGC technicians'3.1.2-6 Amendment No Mg/

SHNPP FSAR a) Licehsed Operators - The licensed operators are responsible for performing shift operations in accordance with the procedures, instructions, .

set points, limitations, and precautions contained in the Plant Operating Manual and the Technical Specifications. They exercise continuous monitoring of plant conditions and system parameters. They manipulate the controls and equipment to start up, change output, and shut down the plant as required by operating schedules and load demands'hey initiate the immediate actions necessary to maintain the plant in a safe shutdown condition during abnormal and emergency situations'hey maintain required records of plant data, shift events, and performance checks. They initiate plant corrective maintenance to report and document equipment problems. Licensed Senior Control Operators (SROs) have the responsibility and authority to assume the control room command function during the temporary absence of the Shift Foreman. The licensed operators report to the Shift Foreman.

b) Non"Licensed Operators - The non-licensed auxiliary operators are responsible to the Shift Foreman for assisting in the performance of assignments associated with shift operations or refueling. The non"licensed operators'uties are normally associated with the operation of auxiliary systems and equipment o'utside the control room. Non-licensed radwaste operators perform shift operations of the Waste Processing Systems. Non-routine operations are performed under the direction of a licensed control operator or Shift Foreman. Radwaste Operators report to the Radwaste Shift Foreman.

c) Radwaste Supervisor - The Radwaste Supervisor supervises the shift operati6ns of the Waste Processing System. This includes the working procedures for the maintenance and implementation of the waste process equipment, and the operation of the equipment necessary to generate process water utilized within plant systems. The Supervisor xs

~ ~ the responsible for ensuring safe and efficient handling and storage of plant-generated contaminated wastes until final disposition. He is assisted by the Radwaste Shift Foremen, Radwaste Operators, Project Specialist - Radwaste, Engineers, and Radwaste Auxiliary Operators. The Radwaste Operations Supervisor reports to the Manager - Operations ~eai-h.

Co~%re )

d) Shift Foremen Radwaste - The Shift Foremen Radwaste ensure the safe, dependable, and efficient operation of the Waste Processing System. It is the responsibility and authority of the Shift Foremen Radwaste to direct the activities of the Radwaste Operators to ensure efficient handling, processing, storage, and shipment of plant generated contaminated wastes. They are supported by and supervise Radwaste Control Operators and Radwaste Auxiliary operators. The Shift Foremen-Radwaste functionally report to the Radwaste Supervisor but are under the direction of the Shift Foreman to ensure that radwaste operations support is compatible with overall plant operations.

e) Operations Support Supervisor - The Operations Support Supervisor provides technical and engineering support to the plant operating personnel.

He is responsible for the implementation and efficient operation of the shift technical advisor (STA) program at the plant as well as for providing direct technical support in the areas of: (1) Plant Operations, and (2) Fire 13.1.2-7 Amendment No. ~ ZP

0 SHNPP FSAR Oper~%<<~ E~~neer~

Protection as necessary to support safe, efficient, reliable operations. He is assisted by ShiEt Technical Advisors, and a ire Protection Sy~kist. The Operations Support Supervisor reports to the Manager Operations.

I ~w<o< Spc'c.I+3 4 j Protection - The Senior Specialist - Fire f) Senior Specialist Fire Protection is responsible for fire detection equipment, Eire, protection equipment, and general safe working conditions for employees. He is responsible for keeping current on "Fire Protection Guidelines for nuclear power plants," Regulatory Guide 1.120, and Branch Technical Position APCSB 9.5-1 and 9.5-1 Appendix A, and informing plant management of changes affecting the plant. He will evaluate damage to plant fire protection equipment under warranty and make recommendations as to course of action. He will coordinate plant inspections for insurance purposes. He is assisted by a S'pecialist and Fire Protection Technical Aides. The Senior Specialist Fire Protection'reports to the Operations Support Supervisor.

g) Shift Technical Advisor The Shift, Technical Advisor provides accident assessment and technical advice concerning plant safety to shift operations personnel. He performs 10 CFR 21 evaluations for the shift operations personnel'e accomplishes this by performing engineering evaluations of off-plant operations, maintaining and broadening his knowledge of normal and normal operations, and diagnosing off-normal events'he Shift Technical Advisors report to the Operations Support Supervisor.

13.1.2.2.3.4 Technical Support Unit The Technical Support Unit provides engineering support for the entire plant staff. 'heir support involves investigations of day-to-day equipment and system operation. Based on their investigations, they recommend modification tasks to keep the plant in co'mpliance with new regulations or to improve efficiency of operation. V'ke fP cz -7eckvuccd gulp H p.js,c, gnat~~ gg Pokey I9z macon 7ce+Frcq~.

The Manager Technical Support Unit develops and tests maintenance modifications and provides technical support for plant 'outages, plant operation, and maintenance and manages the plant Inservice Inspection (ISI) and performance programs. The Manager Technical Support has the authority to issue procedures, Standing Orders, and Special Orders. He is supported by the Engineering Supervisors and a grincipal Engineer. The Manager - Technical Support Unit reports to the Assistant Plant General Manager.

The Engineering Supervisors and a Principal Engineer are responsible for providing technical direction and coordination for plant engineering

'tudies. They develop and implement the inservice inspection program and plant performance programs as well as procedures, instructions, and guidelines for plant engineering functioris. They are supported in these tasks by a staff oE engineers, specialists, engineering technicians, and draftsmen. The Principal Engineer and the Engineering Supervisors report to the Manager Technical Support.

13.1.2.2.4 Startup and Test Unit The Manager - Startup and Test and accomplishing, on schedule, is responsible for successfully implementing the Harris Nuclear Project preoperational ~

13.1.2-8 Amendment No. ~ g7 .

SILVPP FSAR v Harris Plant Operations Startup and Test Unit reports to the General Manager Section.

The Manager Startup and Test is responsible for the following'.

a) Supervises the activities of the Startup Organization through the Startup Supervisors.

b) Prepares and updates the startup schedule.

c) Assigns overall test responsibility to the Startup Supervisors.

d) Reviews and approves requests for vendor assistance as recommended by the Startup Organization.

e) Reviews and approves/recommends approval of test procedures, test procedure modifications, and test data in accordance with the Startup Manual instructions.

f) Reviews and recommends approval of startup requests for construction and engineering modifications or changes required during the test program.

g) Issues periodic progress reports and work schedules for the Startup Organization.

h) 'ssues special reports concerning startup activities as he deems necessary.

i) Reviews progress of startup activities with contractors, vendors, and Company management.

j) Maintains liaison with the plant management, keeping them informed of the test program status, and coordinates with them the activities of plant personnel assigned to startup activities in conjunction with their trainihg program.

k) Represents the Startu'p Organization on interdepartmental and interorganizational committees associated with the test program.

1) 'aintains liaison with contractors and vendors to coordinate their activities relating to the test program.

m) Is responsible for the preparation and ma'intenance of the Startup Manual.

n) Accepts release for tests from Harris Plant Construction Section.

He is supported in the accomplishment of these tasks by a staff of Startup Supervisors,gngineers, specialists, technicians, and clerks. The Manager-Startupgreports to the General Manager - Harris Plant.

cond Test 13.1.2-9 Amendment No. W ZF

SHNPP FSAR The Startup Supervisors are responsible for checking out and starting up on schedule the systems assigned in their areas in accordance with the Startup Manual and regulatory requirements. Each supervisor is assigned engineers and technicians and reports to the Manager Startup and Test Unit.

13.1.2.2.5 Regulatory Compliance The Regulatory Compliance for regulatory compliance activities and routine reporting of ~

Unit provides staff functions to the 'entire plant noncompliance items. The Snit is responsible for the continual updating of the FSAR and Technical Specifications, and it serves as the on-site contact for the NRC.

The Director - Regulatory Compliance coordinates activities at the plant to ensure that commitments, responses, records, and.'eports are prepared, submitted, and maintained in accordance with regulatory requirements. He also maintains a tracking system for the resolution of all'lant safety and environmental concerns. He serves as the on-site contact with NRC and provides the expertise necessary to support plant activities in accordance with the. operating license and technical specifications. He is assisted by a staff of technicians and specialists'he Director - Regulatory Compliance reports to the General Manager Harris Plant Operations Section.

13.1.2-10 Amendment No. X Z7

h SHNPP FSAR ASSIGNMENT OF ON-SITE SHIFT OPERATIONS The Operating Supervisor is responsible for~operating activities at the plant ~ The shift complement consists of one Shift Foreman (SRO), one Senior Control Operator (SRO), two Control Operators (RO), four Auxiliary Operators, and at least one Radiation Control Technician qualified in radiation protection measures. Each shift will also have personnel fulfilling roles in Fire Protection and Radwaste Control (normally five) ~ The 'Harris Plant Operations Section will have shift operating crews assigned to provide 24-hour coverage of plant activities'ach shift operating crew will be manned in accordance with Technical Specification Table 6.2-1. Additional support, for example the 1&C Technicians, Mechanics, Chemistry Technicians, and Plant Storekeepers, will be available on a normal .two shift basis, but this schedule will be subject to change as plant conditions require. On-call personnel will Engineers will also be available as required, although they will normally work a regular schedule.

A During fuel movement operations or core alterations there will be one Senior Reactor Operator in Reactor Containment and an operator in the Fuel Handling Building. This Senior Reactor Operator will direct'and supervise the operation and will report to the Shift Foreman.

The following chart contains the minimum shift assignments of the Operation Unit:

k MINIMUM SHIFT CREW COMPOSITION LICENSE APPLICABLE CATEGORX OPERATIONAL MODES 1, 2, 3, 4 5 & 6 SRO 2 1*

RO 2 1 Non-Licensed 2 ~

.. 1 Shift crew composition, including a Radiation Control technician qualified in radiation protection procedures, may be less than the minimum requirements for a period of time not to exceed 2,'hours in order to accommodate unexpected absence of on"duty. shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements'n the unlikely event an unexpected absence occurs that would involve the health physics technician on 'duty, it is possible this position would be covered by the individual qualified in radiation protection procedures for short periods of time, e.g., a few hours.

Operational Modes listed above are defined in the Technical Specifications. It is ~

expected that the number of personnel as outlined in Table 13.1.2-1 will be'sed to support the operation of, the plant. In the event that additional health physics personnel are required, it is projected that contract health physics services will be used. The number of contract health physics personnel re'quired and their ANSI qualifications will be situationally dependent.

• Does not include the licensed Senior Reactor Operator or Senior Reactor Operator limited to Fuel Handling, supervising core alterations.

13.1.2-11 Amendment No. W Z7

SlVNPP FSAR TABLE 13.1.2-1 (Cont'd)

PROJECTED SHNPP STAFF LOADING TITLE NO ~ OF POSITIONS STARTUP & TEST Manager - Startup & Test 1 Startup Supervisor 5 Startup Engineer 29 Engineering Technician 1 8 Senior Clerks 4 ENVIRONMENTAL & RADIATION CONTROL Manager Environmental & Radiation Control 1 Supervisor Environmental & Chemistry 1 Environmental & Chemistry Foreman Environmental & Chemistry Technician Project Specialist Environmental & Chemistry 2

13 1

l~

Senior Specialist Environmental & Chemistry 2 4x Supervisor - Radiation Control 1 Project Specialist - Radiation Control 1 Senior Specialist Radiation Control 2 Senior Specialist ALARA 1 Traveling Radiation Control Fore'man 1 Radiation Control Foreman 3 Radiation Control Technician Senior Clerk 23 1

I<

Traveling Radiation Control Technician 7 4+

TECHNICAL SUPPORT Manager - Technical Support 1 Engineer Supervisor 2 Principal Engineer 1 Project Engineer 6 Engineer 19 Senior Specialist 9 Co-Op Engineer 3 Co-Op Technician 1 Engineering Technician 1 ll2 Senior Draftsman 13.1.2-14 Amendment No. ~ Z'P

SllNPP F SAR 13.1.3 QUALIFICATION REQUIREMENTS FOR PLANT PERSONNEL*

13.1.3.1 Minimum ualifications Minimum qualifications for plant personnel are listed in CP&L's position on Regulatory Guide 1.8 in Section 1.8.

13.1.3e2 ualification of Plant Personnel P ev-conge1 Resumes through Joe~

Eosjpleuu 13.1.3-35.

( are provided in Tables 13.1.3-1 37

• Further information is contained in the TMI Appendix.

13.1.3-1 Amendment No. ~ 3'7

SHNPP FSAR QE LEYTE TABLE 13.1 ~ 3-1 James L. Millis General Manager - Harris Plant OperationsSection I. Educ tion and Trainin A. B. . Degree in ElectricaL Engineering - 1955 - U. S. aval Academy

- An polis, MD 1958 B. Navy Nuc ear Power'School A. June 1951 - June 79 - U. S. Navy B. August 1979 - Septemb r 1980 Proj t Manager, System Development Corporation Santa Moni , CA September 1980 September 198 Manager, Nucl'ear Training-Southern California'dison pany D. October 1981 - employed Mana r ".

Plant Operations in the Nuclear Operations Department, arris Pla Section. Located in the'eneral Office. 0 E. April 1982 Gen al Manager " Harris ant in the Nuclear Operations Depa tment located at the Har 's Plant, New Hill, NC F. September 19 3 - General Manager " Harris Pl t in Harris Nuclear New Hill., NC

'Project D apartment located at the Harris Plant III'rofessi al Societies A. ~ erican Nuclear Society B N. C. Society of Engineers

13. l. 3-2 Amendment No. 27

1 SIINPP FSAR QEt FYF TABLE 13.1.3-5 Lloyd R. Hancock Administrative Supervisor I. E ucation and Trainin A. ssociate'in Applied Science Degree Mechanical Technol y from vidson County Community College 1971 B. Dipl ma in Mechanical Engineering from the Internati - nal Corre ondence School (ICS), Scranton, Pennsylvani 1974 ZI. ~Ex erience A. September 197 through June 1971 Draftsma , Croft, Inc.

B. . June 1971 employ as an Engineering Ai I in the Power Plant Engineering. Depart ent in the General fice C June 1972 em p lo y ed as a Technician in the Power Plant Engineering Department in the Gene 1 Office D. October 1973 employed as a Tec ician I in the Power Plant Engineering Department in t General Office E. June 1974 employed as a Ju ior ngineer in the Power Plant Engineering Department 1 cated i the General Office and at the Brunswick Plant, South rt, North arolina e

F. October 1975 employ as an Engineer 'n the Power Plant Engineering

,Department located xn the General Offi e and at the Brunswick Plant, Southport, North arolina G. February 1977'ployed as an Engineer in th Generation Services Section of t e Generation Department in the neral Office H. January 79 employed as a Senior Engineer in t Generation Service Section of the G'eneration Department in he General Office I. May 79 employed as a Senior Engineer in the Genera ion Services-0& Section of the Generation Department in the Gener Office une 1979 employed as a Senior Engineer in the Nuclear G eration Section of the Generation Department in the General Office November 1981 employed as an Administrative Supervisor in the Nuclear Operations Department located at the Harris Plant, New ill, North Carolina 13.1.3-9 Amendment No. 27

SHNPP FSAR QF LE7E TABLE 13.1.3-5 (continued)

Lloyd R. Hancock Administrative Supervisor L. September 1983 employe dministrative Supervisor in the Harris Nuclear Project pa nt, Harris Plant, New Hill, NC III. Professional cxeties A. ember of American Society of Mechanical Engineers

13. 1. 3-10 Amendment t1o.+ Sr

SHNPP FSAR TABLE 13.1.3-7

/

Dean L. Tibbitts Project Specialist - Regulatory Compliance I. Education and Trainin A. Webb High School, Reedsburg, Wisconsin - 1969 B. University of Missouri"Rolla, Rolla, Missouri - 1975 - B.S. and M.S.

in Nuclear Engineering C. University of Maryland, College Park, Maryland - 1978 II. Ex erience Prior to Joinin CphL A. June 1975 to March 1980 - U. S. Nuclear Regulatory Commission Project Engineer - Washington, DC B. March 1980 to February 1981 << NVTECH Senior Consultant, Bethesda, Maryland C. February 1981 - March 1983 Phoenix Power Services Principal Washington, DC III Ex erience with CPGL A. June 1983 Employed as a Senior Specialist - Regulatory Compliance in the Nuclear Operations Department at the Shearon Harris Nuclear Power Plant located in New Hill, North Carolina B. September 1983 " Reorganization Department renamed Harris Nuclear

~r ect Depar'tment C. August 1985 Promoted to Project Specialist Regulatory Compliance in the Harris Plant Section'of the Harris Nuclear Project Department IV; Professional Affiliations and Achievements None 13.1.3-13 Amendment No. ~ Q~p

SII:TPP FSAR TABLE 13.1.3-23 (m 0

Danny G. Batten Shift Foreman far I. Education and Trainin A. Bladenboro High School, Bladenboro, North Carolina 1965 B. U. S. Navy

l. Electricians's Mate "A" School - 4 months

2. Nuclear Power School 6 months

3. Nuclear Prototype 6 months II. Ex erience Prior to Joinin CPTtL A. April 1967 to January 1970 - U.S.S. Truxtun DLGN-3S. Qualified Electrical Operator, Auxiliary Electrician and Reactor Plant Shutdown Watch. Maintained electrical equipment.

B. February 1970 to May 1971 - Monob YAG-61 'n electrical s y stem on board this research vessel.

charge of maintaining III. Ex erience with CPSL A. July 1971 employed as Auxiliary Operator "A" in the Generation and System Operations Department at the H. B. Robinson Plant, Hartsville, South Carolina B. November 1972 employed as Control Operator in the Bulk Power Supply Department at the H. B. Robinson Plant, Hartsville, South Carolina C~ August 1977 employed as Senior Control Operator in the Generation Department at the H. B. Robinson Plant, Hartsville, South Carolina D. June 1981 employed as Shift Foreman Nuclear in the Nuclear Operations Department at the H. B. Robinson Plant, Hartsville, South Carolina E~ May 1982 employed as Shift Foreman Nuclear in the Nuclear Operations Department at'he Shearon Harris Nuclear Power Plant, New Hill, North Carolina. (Temporarily assigned to the H. B. Robinson Plant, Hartsville, South Carolina.)

F. July 1982 employed as a Shift Foreman in the Nuclear Operations Department at the Shearon Harris Nuclear Po~er Plant, New Hill, North Carolina G. September 1983 employed as a Shift Foreman in the Harris Nuclear Project Department, New Hill, NC

13. 1~ 3-37 Amendment No.

Sli:<PP FSAR TABLE 13.1.3-25 John W. Digby Shift Foreman I. Educ ion and Trainin A. Mi i Edison Senior High Miami, Florida 1960 B. George T. Baker Aviation - Miami, Florida - 1964 t 1966 - No Degree Avaiation Mechanics C. Purdue - W t Lafayett, Indiana - 1980 to 198 No Degree - STA Program D. Electrical Powe Production Technical S eppard Air Force Base, Watchta Fall, Te s October 1961 to J e 1961 Electrical Power Production E. Electric Power Produc on'Missile hool " Sheppard Air Force, Watchta Fall, Texas - ne 1961 t November 1961 Electrical Power Production II 'x erience Prior to Joinin CPS A. June 1961 tb March 1964 EPPT United States Air Force B~ June 1965 to Septemb 1966 - Truck river - Lou-Mack Transfer-Miami, Florida C. September 1966 June 1978 Watch E'ngi eer (SRO) - Florida Power and Light Comp ny Miami, Florida D. June 1978 t June 1980 - Reactor Control Ope tor I- Washington Public Po er Supply System Richland, Washing on E. August 980 to January 1984 - Shift Supervisor - ublic Service Indi a - New Washington, Indiana III. Ex er'ence with CPSL ~

A February 1984 " Employed as Shift Foreman Nuclear in the arris Nuclear Project Depart;ment, New Hill, North Carolina 13.1.3-39 Amendment No. K'7

SHNPP FSAR TABLE 13.1.3-32 LE P. Capps Superintendent Materials and Custodial I. Educ ion A. Pem roke State University " Pembroke, North Carolina 1973 B. S. in Acco ting I A. 1973 Junior ccountant; Plant Accountin Section, Treasury and Accounting Depa tment, Carolina Power & xght Company, Raleigh, North Carolina B. 1975 - Accountant, P nt Accountin Section, Treasury and,Accounting Department, Carolina P er & Lig Company, Raleigh, North Carolina Co 1975 Accountant, Constr ti Engineering and Accounting Section, Power Plant Construction De rtment, Carolina Power & Light Company, Raleigh, North Carolina Do 1976 - Senior Accounta g, Nuclea Construction Section, Power Plant Construction Departm t, Carolina wer & Light Company, Shearon Harris Nuclear Pow Plant, New Hill, North Carolina ~

E. 1978 >> Supervis r - Project Accounting, arris Site Management Section, Nucl ar Plant Construction Depar ent, Carolina Power &

Light Compa , Shearon Harris Nuclear Power lant, New Hill, North Carolina F. 1984 - Superintendent - Materials and Custodial oject in the Admi xstration Section, Harris Nuclear Project Dep tment, Carolina Po r & Light Company, Shearon Harris Nuclear Power lant, New Hill, rth Carolina.

III. ofessional Societies None 87 13.1.3-48 Amendment No. ~

TABLE 13.1.3-33 E. E. Johnson Principal Specialist Document Services I. Educe on A. Nort Carolina State University - Raleigh, North Ca olina - 1964 B. S. 'n Applied Mathematics II. Ex erience Prior to Joinin CP&L A. 1964 - Enginee in Atomic Power Division Newport News Shipbuilding

& Dry Dock Compa B. 1971 - Engineering pervisor in At mic Power Division, Newport News Shipbuilding & Dry Do Company.

C. 1973 " Structural Design ec'ist, Gilbert Associates, Inc.

D. 1975 Project Control En 'n r - Gilbert Associates, Inc.

E. 1977 Records Managem t Consul ant of VEPCO's North Anna Nuclear Power Plant for Gilb t Associate Inc.

1979 Supervisor Administration and ecords Management, at VEPCO's North Anna Nucl ar Power Plant G. 1981 Info ation Management Consultant a SCE&G's V. C. Summer Nuclear Po er Plant for Gilbert Associates, c.

III. Ex erienc with CP&L A. 83 Project Specialist Nuclear Operations Adminx tration, Carolina Power & Light Company 1984 - Principal Specialist - Document Services in the Pro ect Administration Section of the Harris Nuclear Project Depart nt, Carolina Power & Light Company 13.1.3-49 Amendment No ~ Z7

S1KPP FSAR TABLE 13.1.3-33 (continued)

E. E. Johnson Principal Specialist Document Services III. Professional Societ s A. Member of Institute o Certified R ords Managers B. Member of Association of In mation and Image Management (formerly National Micrographics ociat') U pg C. Member of Associ 'on of Records Manag s and Administrators P(

D. Member of clear Records Management and Adm 'strators

' l7(

(1) ember on Micrographics Committee Member on Technical Support Center/Emergency Offs e. Facility Records Committee 13.1.3-50 Amendment No.. PK Z7

SHNPP FSAR TABLE 13.1.3-34 Ronald E. Gurganus roject Engineer Project Analysis, Harris Nuclear Project Depart nt I. Educ tion A. No h Carolina State University - Raleigh, North Car ina - 1970 B.S. in Forest Management.

B~ Duke Un ersity - Durham, North Carolina - 1980 - MBA in Business Administr tion 1982 C. Internationa Correspondence School Scr ton, Pennsylvania Diploma in Ele trical Engineering, Pow'er Option D. 3286th United Sta s Army Reserve Sc ol Raleigh, North Carolina-1982 - First Sergea t School " Dip ma II. Ex erience Prior to Joinin &L A. 1970 " Assistant to Vice P ident of General Construction, Stackhouse, Inc..

III. Ex

~

~

~ erience with

~ ~

CP&L A. 1973 - Transmis 'on Line Right"Of-Way orester, CP&L System Operations Dep rtment Bo 1978 - Sen' Specialist, CP&L System Oper ions Department C ~ 1983 - enior Engineer - Project Analysis, CP Harris Site Manag ent D. 19 4 - Project Engineer " Project Analysis, Harris roject I

dministration, CP&L, Harris Nuclear Project Departm t IV. rofessional Societies None

13. 1. 3-51 Amendment No. ~ g7

SHNPP FSAR TABLE 13.1.3"35 R. E. Butler Project Engineer - Site Industrial Engineering I. Educat'on A. Mary nd University College Park, Maryland 1959 ~ S. in Chemic 1 Engineering/Math B. Maryland iversity College Park, Maryland '67 B. S. in Industrial gineering II. Ex erience Prior to J 'nin CP&L A. 1964 - Distribution upervisor " UP - Landover, Maryland B. 1966 Personnel Supervi or U Landover, Maryland C. 1967 - I.E. Manager " UPS - ew Jersey District D. 1974 Industrial Engine ring nager Assigned to UPS National Staff New York, New ork E, 1976 District Op ations/I.E. Mana er Carolina District - UPS-Raleigh, North C olina III. Ex erience with 6L A.. 1980 dustrial Engineering Supervisor - Har 's Site Construction B. 1983 - Project Engineer " Site Industrial Engineer ng, Shearon H ris Nuclear Power Plant IV. P fessional Societies A. Member American Institute of Industrial Engineers 13.1.3-52 Amendment No. PCZ7

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IFT FOREMAN STA'S SHIFT FOREMAN SKI CR RAG I AT ION CCNTRCL J RW CCNTRK. CP ERAT I CNS C~CL SUPERV I SCR I ~ACTORS ENG I NEERS CFERATORS SRO PROJECT F I RE SPECIAL I ST I AUXIL I ARY PROTECT I CII RAG I AT I CN I GPERATGRS ~CI AL I STS FiT)jCd AUXI L I ARY LEGENO Egn~T'dupccsVe CPGVTCRS AOMINISTRATIVE ORGAIIIZAT,ION SRO SeHIOR REACTOR OPERATORS LlCENSE RO REACTOR OPERATORS LlCENSE Z7 AMENDMENT NO. 2K SHEARON HARRIS NUCLEAR POWER PLANT Carolina Power 5 Light Company FINAL SAFETY ANALYSIS REPORT FACILITY ORGAt4IZATION FIGURE 13.1.2-2 l." ~ 1g QJ 'su ag I T I SHNPP FSAR 13.~ 2.~ 2 ~ REPLACEHENT AND RETRAINING A training program will be. utilized to maintain the proficiency of the plant operating organization after the initial plant start-up. This training program will include, as described below, requalification training for licensed personnel, and replacement training for replacement personnel. 13.2.2.1 Licensed 0 erator Re ualification Trainin Qvak uca re op~ c'A Following the initial licensing of cold license can idates, a;requalification training program will be initiated to ua&~a and demonstrate the continued prosaic<<'~cq a~&competence of all licensed personnel. This requalification training program %he will include pre-planned lectures, on-the-job training, and regular and continuing operator evaluation. The SHNPP simulator will be used to fulfilla ro fate rtions of this retraining program.

13. 2.2. 1. 1'ec tures w'i+h tocia55. 'Requc.(lfi'c~y,'~~ +rc \pJl&g ~sO Kcc+a ~Q GAL lo CPR 5'5 peqapgrc~c'A S oscP ~ +wp ye~ +~~'+ J A minimum of six pre-planned lectures will be presente urging eac requalification cycle. These lectures will be scheduled throughout thenyearcgcha taking into account heavy vacation periods and infrequent operations such as refueling periods and forced outages Lectures may be deferred due to unanticipated shutdowns. However, these lectures shall be conducted as soon as practicable thereafter. Content of the'lectures shall take into

.consideration the categories as listed in 10CFR Part 55, Appendix A, heat transfer, fluid flow, thermodynamics, mitigation of accidents involving a degraded core, operating experiences from similar plants and the results of the annual examination. Training aids such as fil'ms, video tapes, and slides may be used and some self-study may be required in conjunction with the lectures. An instx'uctor will present or attend as an auditor at least 50 percent of the lecture series. All licensed individuals will be required to attend every pre-planned lecture except those specifically exempted. Exemptions will be allowed only for individuals scoring greater than 80 percent in the corresponding area on the previous examination.

13. 2. 2. 1. 2 On-the-job Training The on-the-job training portion of the requalification program will consist of the following:

a) - Control Hanipulation -Licensed reactor operators shall manipulate and senior reactor opexatoxs shall manipulate or direct or evaluate the activities of those manipulating the station controls through a minimum of ten reactivity changes during each annual cycle. These manipulations may consist of any of the following, providing that asterisked items are performed annually and all other i,tems are performed on a two year cycle'.

• 1) Start-up to the point of adding heat

2) Orderly shutdown

13. 2. 2-1

SHNPP FSAR of auxiliary Eeedwater flow due to loss of off-site power, loss of main feedwater ELow, safety injection signal, or low-Low level in one or more steam generators.

6) Visually observe system piping and components during system operation for abnormal vibration or piping response to system and component operations. Perform instrumented tests as required by the System Dynamic Test and Analysis Test Procedures.

ing Low Power Testing, prior to attaining 25X power rm a 48-hour nce test of each motor-driven 'ary Eeedwater pump using a flow 0 the stea rators and recirculation flow to Condensate Stora On completion, each pump will remain idle a 'tely eight hours. . it. pump.cooldown, and t will be restar ted and operated using me flow ath Eor one hour. Acceptance Criteria Automatic initiation of feedwater flow Erom the Auxiliary Feedwater System shall occur within one minute of the automatic auxiliary feedwater actuation signal listed in FSAR Section 10.4.9.2.4 (a) and (b).

2) ~ The turbine-driven auxiliary feedwater pump and turbine performance shall meet or exceed vendor performance data supplied in SHNPP Technical Manual 16-P043-3065..

r

3) The turbine-driven auxiliary feedwater pump, reliability"shall be demonstrated by performance oE five consecutive cold starts.

The motor driven auxiliary Eeedwater pumps pressure controL valves maintain pump discharge pressure above 1000 psig.

5) The motor-driven auxiliary Eeedwater pumps fLow to the steam generators shall meet or exceed the rate shown on the vendor pump curves. 'This rate shaLL be equal to or greater than 400 gpm per pump as indicated by installed flow elements.

Note'Steam Generator capacity in gaLLons/inches determined during initial steam generator filling.

6) System dynamic testing completed per SHNPP Dynamic Test and Analysis, FSAR Section 3.9.2.

7) The motor driven auxiliary feedwater pump shall not exceed the limitations for vibration, bearing, and bearing oiL outLet temperatures as specified in SHNPP TechnicaL Manual 16-P043-3065. In addition, the RAB Ventilation and Equipment Cooling Systems shall maintain environmental conditions of the Auxiliary Feedwater piping area within the design requirements of FSAR Sections 9.4.3 and 9.4.5.

E7 14.2.12-37 Amendment No. SHNPP FSAR d) Acceptance Criteria

1) The ESFAS components actuate into the states shown on FSAR Tables 7.3.1-5 through 7.3.1-11 as appropriate for A train operation, 8 train operation, and both train operation. During single train operation, the opposite train 6.9 kV buses remain de-energized.

2) The emergency diesel generators start and sequence loads, including capability to carry manual loads of FSAR Table 8.3.1-2 when offsite power is not available.

3) Upon resetting the initiating FSFAS signals, the safety related

, components actuated above remain in their 8M et q cP<g Mod.e. 14.2.12.1.60 Process Computer Test Summary a) Test Objective To demonstrate that the ERFIS computer system functions as per vendor's technical manual to provide monitoring, alarming, displaying,'eporting, and archiving capabilities to the Control Room Operator, the Technical Support Center, and the Emergency Operations Facility. b) Prerequisites

1) The general prerequisites are met.

2) Hardware is complete and verified by the vendor test plan and procedures.

3) Software is installed and proven by satisfactory completion of the performance sections of the Vendor Test Plan and Procedures.

c) Test Method

1) Verify each process input is operable by simulating the input or application of a known input.

2) Verify that the alarm and conversion of each type process input provides valid information by simulating various input of conditions or by monitoring various levels of known inputs.

3) Verify the ERFIS system's capability to give proper computational results by simulating the inputs or using static test cases and comparing the result against independently computed values.

4) Verify the system display capability in the Control Room, the Technical Support Center (TSC) and in the Emergency Operations Facility (EOF).

5) Verify the redundancy capability of the ERFIS system by inducing system faults and observing system performance. ~

I

14. 2. 12-62 Amendment No. ~37

SHNPP FSAR 1 \

3) With continuous sampling flow established, verify the operation of the temperature control system.

4) Test the operation of the Hotwell Sample Pumps.

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5) Operate the Steam Generator- Blowdown sample isolation valves and verify closure times specified by FSAR Section 6.2.4.

Acceptance Criteria ~ . iir ~ ~ k i- ~ -C') Sample points have been verified per FSAR Table 9.3.2-1 and 9 '.2-2

2) All controls and alarms function in accordance with latest design documents.

3) ~tem~Vemperature control system has demonstrated lii: ~

the 37 capability to. maintain sample temperatures. at 77 + 5 F. \

4) Steam Generator Blowdown samples are cooled to less than 120 F.

5) Hotwell Sample Pumps operate in accordance with vendor instruction manual (16-P175).

~ ~

6) Steam Generator Blowdown sample isolation valves have demonstrated closure times specified by FSAR Section 6.2.4.

14 ' '2.1.79 Loss of Instrument Air Test Summary a) Test Objectives

1) .'o demonstrate that pressure causes fail-safe a reduction and loss of instrument'ir operation of pneumatically-opeiated valves and dampers both safety and nonsafety related located in "the reactor building, auxiliary building and fuel handling building.

') Prerequisites ~w e

1) The general prerequisites are met.

2) Specific prerequisites will be delineated in the system preoperational test procedure.

c) Test Method 4 ~

1) Where safe to personnel and equipment; a slow reduction in pressure and a loss of pressure test will be performed; Testing will be done in small segments/individually and response noted for both safety and nonsafety-related valves and dampers. The loss of pressure test will be conducted by isolating segments/individual items and venting the air from the isolated segment.

Acceptance Criteria

1) Proper fail-safe operation of valves and dampers subject tD a reduction and loss of instrument air is verified.

37 14.2.12-77 Amendment No. ~ . SHNPP FSAR

3) Repeat steps 1 and 2 for low pressure heaters g3 and 4.

Acceptance Criteria

1) .With the bypassing of either heater group, the resultant transient will be less severe than the predicted response of FSAR Section 15. 1.1.

2) The most severe loss of feedwater heater transient will be identified based upon the resultant data from the series of tests.

14.2.12.2.30 Main Steam Isolation Valve Test Summary a) Test Objective

1) This test will demonstrate the capability of the main steam isolation valve (MSIV) "Test Feature" to operate as designed under maximum steam flow conditions.

b)'rerequisites

1) Plant is at approximately 100X power.

2) The general prerequisties are met.

c) Test Methods 1). 'The MSIV control switch is put in the "Test" mode. (Each valve is tested separately)

2) Monitor plant response and valve stem travel.

d) Acceptance Criteria MSIV stem travels from 100X open to 90X open and back to 100X open, 37

2) The plant's control systems operate to maintain steady state conditions.

14.2.12.2.31 Steam Generator Test 'for'ondensation Induced Water Hammer a) Test Objective

1) To demonstrate the capability of transferring feedwater flow from the auxiliary feedwater nozzles to the main feedwater nozzles.

b) Prerequisites

1) Plant conditions are established as required by test instructions for 15 percent power level.

37 14.2.12-99 Amendment No.~ SHNPP. FSAR c) Test Method

1) Operate the plant at approximately 15 percent of full power by feeding the steam generators through the auxiliary feedwater nozzles.

2) Transfer the feedwater flow to the main feedwater nozzles by opening the main feedwater isolation valves per general plant operating procedures.

3) Observe and record the transient that follows.

d) Acceptance Criteria Either low amplitude or no condensate-induced water hammer is observed in the region of the main feedwater nozzle/preheater section of steam generator. 3 ~o4r dri~c~ 14.2.12.2.32 Steam Turbine-Driven Auxiliary Feedwater PumpsEndurance Test PSALM a) Test Objective

1) To demonstrate the capability of the steam turbine-driven c ~b ~e4or dr'i~a~ auxiliary feedwater pumps to continuously feed two or more steam generators Eor a 48-hour periods ~,'g,<<q ~~~e~g; cdmcl a4iou 5 oW Scc4sor 9.g.$ ~J g.q.5 b) Prerequisites

1) Theneral prerequisites have been met.

2) The plant is operating at a power level of 25 percent or less 4m;

~ sl f I 4, ~os~ Tee 'L Jwo+cI Pv~p lw opera4<oA ~ ~ Test Method W~ser+ """ QI OMQCLQC 1 With the steam turbine"driven auxiliary Eeedwater pump supplyin I4.2.LZ -looc, Eeedw er to two or more steam generators, increase reactor power ntil maximum w from the pump is obtained or reactor power is 25 rcent.

2) Maintain ma um flow condition or 25 percent rea or power Eor 48 hours.

3) On completion of 48-ho endurance ru shift feedwater supply to the motor-driven auxiliary Eee ter a secure the turbine-driven auxiliary feedwater pump per norma ant procedure. Additional feedwater needs during this tim ill provided by the main feedwater pumps.

4) After obtainin mbient pump conditions, sta the turbine-driven auxiliary Eeedwa r pump and operate for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

5) Dur'he 48-hour run of the pump, the pump cubicle hu ity will be m ured at specified intervals and cubicle temperature shall r orded at a minimum frequency of hourly.

3l 14.2.12-100 Amendment No. 94 SHNPP FSAR d) Acceptance Criteria

1) The turbine-driven auxiliary feedwater pump and turbine, vibration, bearing, and bearing lube oil temperature shall not exceed Qo+ P wc 44( +~[ef4441 c >

) 4po g r'col 4 0 i t + cMper04'uf 8 4pJc\ 2 0 M<( 5 V I Qf ~410m) ~ ~ 37

2) C4e momtor - clra~e< 4v>a I imp Wee Jlwca4r pv~p ~odor s p 4 s shall ~+ ~~ceca ~g~ F

+c~pero4mes> Qo P t <s c Rr Motor ~awQ g 4c~ er<4urCz ~ ~ u'Ar~4<<V p.o ~, ls. ~~Sar+ t46-2 L2- too +o >~Be ~

1) Operate the steam turbine"driven and(or the motor-d'riven auxiliary feedwater pumps to supply feedwater to two or more steam generators. The initial total flow from the pumps, with all three pumps in operation, will be approximately 500 to 600 gpm. If temperatures and vibration at this flow are acceptable, the test will continue at this flow rate. If temperatures or vibration are not acceptable, the flow rate will be increased until it is acceptable.

The test for each pump will continue for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> "from the time of initial pump starts During this time, pump operating parameters, . (vibrations, bearing temperature, discharge pressure, etc.) and pump cubicle environmental conditions (temperature on an hourly frequency, humidity at specified intervals) will be recorded.

2) Makeup to the Condensate Storage Tank wilL be via an additional condensate transfer Line downstream of the Condensate Polishing System in order to maintain chemistry.

3) Upon completion of the 48-hour endurance run, maintain feedwater supply through the Main Feedwater Pump and secure all three auxiliary feedwater pumps.

4) After obtaining ambient pump conditions for the turbine-driven pump, restart the turbine-dri.'ven pump and operate for one hour.

5) Upon cooldown of the motor"driven pumps (approximately eight hours after completion of the 48-hour run), restart pumps and operate for one hour.

'37 14.2.12-100a Amendment No. ~ SIILIPP FSAR EX) The Reactor Auxiliary Building Ventilation and Equipment Cooling conditions of the auxiliary feedwater ~~ Systems shall maintain environmental areas within the design requirements of FSAR Sections 9.4.3 and 9.4.5. 14.2.12.2.33 Resistance Temperature Detector (RTD) Bypass Flow Verification Test Summary a) Test Objectives

1) To determine that the flowrate in each RTD bypass loop is sufficient to achieve the design RTD bypass transport time.

b) Prerequisites

1) The reactor is in the hot shutdown'ondition with all reactor coolant pumps running.

2) RTD bypass loop flow instrumentation is calibrated and in service.

3) The general prerequisites are met.

I c) Test Method

1) The flow required to achieve the design reactor coolant transport time is determined by measuring and recording the lengths of installed piping in each hot and cold leg.bypass loop and then calculating the flow necessary to achieve design transport time.

2) Total bypass flow rate for each loop is measured and recorded, and then actual bypass transport time is calculated.

n d) Acceptance Criteria

1) The flow rate in each RTD bypass loop yields design transport time as per the Westinghouse NSSS Startup Manual.

14.2.12.2.34 Secondary Sampling System Test Summary a) Test Objective

1) To verify the proper operation and performance of the Secondary Sampling System.

2) To verify the operation of the following sample points S7, S8, S9, S10, Sll.

b) Prerequisites

1) To the extent practical systems to be sampled are at operating pressure and temperature.

2) Cooling ~ater is available to sample panels for sample coolers and chiller condensers.

'3'7 14.2.12-101 Amendment No. ~ SIQIPP FSAR .) Test Method

1) By valving "in and-out" sample lines, verify samples to sample panels are correctly identified.

2) With continuous sampling flow established, verify the operation of the ~mack+ temperature control system.

d) Acceptance Criteria

1) Sample points have been verified per FSAR Table 9 '.2-2.

2) *ata~natka temperature control system has demonstrated the capability to maintain sample temperatures at 77 + 5 F ~

3'7 14.2.12-102 Amendment No. Pj SHNPP FSAR TABLE 15.6.5-6 RADIOLOGICAL CONSEQUENCES OF A POSTULATED LOSS-OF-COOLANT ACCIDENT Doses (rem) T~ntnid Whole Bod Skin 0-2 Hour Dose at the Exclusion Area Boundary"- 1.5.x 10 2.6 x 10 0-30 Day Dose at the Low Population Zone- 1.7, x 10 1 ~ 6 x. 10 0-30 Day Dose to the Control Room Personnel 5.1 x 10 6 '0/ x 10 le4 x 10

• These doses include contributions from ECCS outside containment (refer to Section 15.6.5.4.3.c for further information) ~

15.6.5-17 Amendment Nn.~ 37 SHNPP FSAR 15o7.4.3.2 Postulated Fuel Handling Accident Inside Containment The possibility of a fuel handling accident inside Containment during refueling is relatively small due to the many physical,'dministrative, an fet restri tions imposed on refueling operations. ~ectuu3ctZ~ ehgCa d CO~IC t~d CWV VeWVLc %tO~ t.ghbia+lm~ MO~t'4Og~. During uel andlxng operations, the Containment is kept in an isolable condition, with all ventilation penetrations to the outside atmosphere either closed or capable of being closed on aa- isolation signal initiated by <<M~ radiation monitors in the Containment'andling ~~ . At least one of the two interlock doors on the personnel locks is kept closed. In addition, there are airborne area during refueling. Should a fuel assembly be dropped and release activity above a prescribed level,- the radiation monitors sound an audible;-= alarm, the Contain&'ent is isolated and, the personnel are evacuated. The containment pre entry-purge lines are automatically closed, thus minimizing the escape of any Vpg W a. Con'tc ~~em% verb t(c&lo~ i~oLc.+io~ +xq&o ~ 15.7.4-2a Amendment 1o. ~ SHNPP FSAR radioactivity. The consequences. of dr'opping a fuel assembly in the Containment are less severe than the consequences of dropping the assembly in the Fuel Handling Building, since the Containment .provides a considerably greater holdup time than the Fuel Handling Building, allowing Eor radioactive decay of the released fission products. For analytical purposes, consideration is given to one accident; a drop of a fuel assembly into the refueling cavity by the manipulator crane inside Containment. Assumptions and parameters used in evaluatin fuel'andlin accident inside Containment are shown in Table 15.7.4-'3 ~ + ~us awu~~g+h~+ 4mf+hcr rc$ g~~ < s ~ cr 4 The were radiological consequences of conservatively evaluated by assuming a occur thp containment releases ~~Athe first ~second ~ fuel handling accide'nt inside Containment period. 37 made in a controlled manner- through the Reactor: .-.;,. Auxiliary Building Filtration System charcoal adsorbers. The assumption of a . controlled release was made due to the availability of these charcoal adsorbers and based on calculations showing that containment isolation can be +<< ~~ +< Se+poWt t-c:buccal +o I50 M/4r or to roc4ovucla'he<'ms'~g ~ g<<~+ co~(e:i~~~t i~o4Ãi'o vc.4c. e acti ity released inside Containment as a result of a fuel handling aqerc"rio~> accident will be detected by ese radiation 7 monitorS. The response time for ebea monitorS is expected to be Less than $ 4& second ~ Following activity <c~<'~"><< closure of the containment ~anhay isolation valves. The valves will 0 require 15 seconds to close. The Containment will be isolated in Pf seconds " a ter detection of the accidental release of radioactivity. The time required for airborne activity to reach the containment isolation ') detection, the monitorswiLL initiate the valve is based on 1) travel time from tPe surface of the reactor pool to the nearest intake header and 2) travel time through the duct. The nearest intake header from the pool is located at a distance of 14.8 Eeet. The average airflow velocity'ithin a distance of Less than 3 feet and the velocity at 3 feet from the intake header was estimated using equations from Reference 15.7.4-1. These equations are as Eollows: v=~ 10X + A v -1 X/lOA tan A (2) X/10A where, X = distance outward along axis, Et. (Note: Equation is accurate only when X is less than 1 1/2 D) V centerline velocity at distance X from hood, ft/min. Vav = average velocity within a distance X, ft/min. Q = air flow, cfm area of hood opening, ft 37 H.7. C-3 Amendment No. ~ SllilPP FSAR D = diameter of round hoods or side of essentially square hoods. The air velocity beyond 3 feet, though expected to be smaller, is assumed to remain the same as that at 3 feet from the intake header. The size of the intake header is 24" x 24". The average air velocity up to and including 3 feet was estimated at 179.6 ft/min and the average air velocity beyond 3 feet was estimated to be for a given intake rate of 2500 cfm through the header. Therefore the activity would take 27.6 seconds to reach the intake header from the surface of the pool. The travel time in the 26 in. x 30 in. duct (106 ft. in length) would be 3.4 seconds. Th'erefore the total time required before the activity would reach the isolation valve is 31 seconds'he following conservative assumptions.,are based on Regulatory Guide- 1,25 ..'. and inherent plant design parameters used to calculate the activity releases and offsite doses for the postulated fu'el handling accident inside Containment. a) The accident is assumed to occur 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following reactor shutdown for refueling. b) All rods in one fuel assembly are ruptured. c) The assembly damaged is assumed the highest powered assembly in the core regio'n to be discharged. The values for individual fission product inventories in the damaged assembly are calculated assuming full power operation'at the end of core life immediately preceding shutdown. A radial peaking factor of 1.65 is used. d) All of the gap activities in the damaged rods is released and consist of the 10 percent of the total noble gases other than krypton-85, 30 percent of the krypton-85, and 10 percent of the total radioactive iodine in the rods at the ti:me of the accident. e) The iodine gap inventory is composed of inorganic species (99.75 percent) and organic species (0.25 percent). f) The refueling cavity water decontamination factors for the inorganic and organic iodine are 133 and 1 respectively, giving an overall effective decontamination factor of 100. .g) The retention of noble gases in the refueling cavity water is negligible. h) The accident occurs during refueling with the Containment Purge System operating. gO i) Containment isolation occurs M seconds after detection of the accident 3'7 with resulting filtration. Ro j) A filter efficiency of + percent is used for halogens. The doses from a fuel handling accident occurring inside Containment have been calculated, and have been found to be below the guidelines of 10CFR100. The results of this analysis are presented in Table 15.7.4-4. A~e~h~e~+ nlrb.S'7 15.7 '-4 .MqogQmo~~ SHNPP FSAR TABLE 15 '.4-3 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSE UENCES OF A FUEL HANDLING ACCIDENT INSIDE CONTAINMENT Design Basis Realistic Parameter Assum tions Assum tions Source Data: Power level, MWt 2900 2775 Radial peaking factor lo65 1.55 Burnup 3 full.-power 3 full-,power years years Decay time, hr 48 48 Number of failed assembly 1 1/17 Fraction of fission product gases contained in the gap region of ~ the fuel rods, percent Kr-85 30 30 Other Noble Gases 10 10 Iodine 10 10 Activity Release Data Fraction of gap activity released to pool 100 100 Minimum water depth above damaged rod's, ft. 23 23 Pool decontamination factor for noble gases Pool decontamination factor for iodine Inorganic '33 Organic 1 Overall 100 500 Iodine chemical form released to fuel building Inorganic iodine percent 75 75 Organic iodine, percent 25 25 Containment Isolation Time 37 Accident (Sec) >e-Cxrg ac5~ Containment/Purge (~) 37,000 37,000 Containment Volume (cu ft.) 2.37 x 10 2.37 x 10 15.7 '-8 Amendment No. 37 SHiVPP FSAR TABLE 15.7.4-3 (Cont'd) Design Basis Realistic Parameter Assum tions Assum tions Filter Efficiency 37 Iodine, inorganic percent ~9o 99 Iodine, organic percent 95 9o 99 Noble gas percent 0 0 Activity released to atmosphere, (ci) Before ~Isoto e Isolation Total* 's I-131 2.8~ 5'Co~ x 10 I-133 5.3~ x . 10 37 Xe-131m 4 ' x 10 Xe-133 1.0 x 10 Xe-135 2 ' x 101 Kr-85 l- l~xs ~ 10 2 ' x 10 Dispersion Data Atmospheric 5 percentile level 50 percentile level dispersion factors X/Qs, (Table 2.3.4-5) X/Qs, (Table 2.3.4-5) Dose Calculation Dose model Hodel as discussed in Appendix 15.0A - Total is the combination of doses before isolation and after through controlled purge. 15.V.4-9 S7 Amendment No. W SHNPP FSAR TABLE 15.7.4-4 RADIOLOGICAL CONSE UENCES OF A POSTULATED FUEL HANDLING ACCIDENT INSIDE CONTAINMENT desi n Basis Assum tions Result Before Isolation Total* Exclusion Area Boundary Dose (0 to 2 hr.) (rem) '37 Thyroid Whole body ~9.5 3.4 V>o x 10 3 lg ~x 4.7 x 101 lo-l LPZ Outer boundary Dose (duration) (rem) '37 Thyroid Z.Z.~ x 10 1 Whole body 7.6 x 10 1.1 x 10 /r

• Total is the combination of doses before isolation and after isolation through controlled purge.

37 15.7.4-10 Amendment No. ~