ML19344D741
| ML19344D741 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 04/09/1980 |
| From: | Baer R Office of Nuclear Reactor Regulation |
| To: | Stampley N MISSISSIPPI POWER & LIGHT CO. |
| References | |
| NUDOCS 8004280084 | |
| Download: ML19344D741 (9) | |
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UNITED STATES y))
(,g NUCLEAR REGULATORY COMMISSION g.* \\
'f6 : C WASHINGTON. O. C. 20555
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APa o a 1980 v
Docket Nos. 50-416 and 50-417 Mr. N. L. Stampley, Vice President Production and Engineering Mississippi Power and Light Company P. O. Box 1640 Jackson, Mississippi 39205
Dear Mr. Stampley:
SUBJECT:
REQUESTS FOR ADDITIONAL INF0PMATION (Grand Gulf Nuclear Station, Units 1 and 2)
As a result of our review of the information contained in the Final Safety Analysis Report for the Grand Gulf Nuclear Station, Units 1 and 2, we have developed the enclosed requests for additional information.
Included are questions from the Instrumentation and Control Branch concerning Sections 7.4 and 7.5.
We request that you amend your Final Safety Analysis Report to reflect your responses to the enclosed requests as soon ts possible, and to inform the Licensing Project Manager, Thomas C. Houghton, of the date by which you intend to respond.
Please contact us if you desire any discuision or clarification of the enclosed requests.
Sincerely,
&& k-
'1 R ert L. Baer, Chief Light Water Reactors Branch No. 2 Division of Project Management
Enclosure:
Requests for Additional Information ces w/ enclosure:
See next page 8 0 0 4 2 8 00fsi
'Mr.N'L.Stampley jgggy9gggg Mr. N. L. Stampley Vice President - Production Mississippi Power and Light Company P. O. Box 1640 Jackson, Mississippi 39205 ccs: Mr. Robert B. McGehee, Attorney Wise, Carter, Child, Steen and Caraway P. O. Box 651 Jackson, Mississippi 39205 Troy B. Conner, Jr., Esq.
Conner, Moore and Corber 1747 Pennsylvania Avenue, N. W.
Washington, D. C.
20006 Mr. Adrian Zaccaria, Project Engineer Grand Gulf Nuclear Station Bechtel Power Corporation Gaithersburg, Maryland 20760 l
IS IO31.0 INSTRUMENTATION AND CONTROL 0031.70 Identify the criteria that restricts consideratien of diversity to (7.4.1.1 3 5) initiating signals and only to the RPS, ECCS and containment isolation systems.
10 CFR 50, Appendix A states that specific requirements for redtndancy and diversity have not been developed er defined but still must be censidered in arriving at a design that satisfies the necessary safety requirements. RCIC is both redundant and diverse with HPCS fcr the safe shutdown function.
0031.71 Apparent inconsistencies and cmissions were noted in the analysis (7.4.2.1) fcr ecmpliance with the following criteria.
Amend the FSAR as (7.4.2.2) required.
A.
- 1) Regulatcry Guide 1.6.
Justify your statement that because the single failure criteria is not applicable, RG-1.6 is not applicable to RCIC.
- 2) General Cesign Criteria 21.
The analysis addresses testability but does not address reliability.
/
- 3) General Cesign Criteria 29 The analysis merely states the function of RCIC; it does not address the probability of the system functioning when needed.
- 4) General Cesign Criteria 34 The analysis consists of a reference to a non-existent subsection.
- 5) IEEE 279 The discussicn presented under paragraph 4.12 is pertinent to paragraph 4.13 and is unrelated to paragr.aph 4.12.-
The discussien should be moaified as needed and relocated to paragraph 4.13 It appears that there are no operating bypasses as defined in IEEE 279
associated with RCIC.
E.
- 1) General Cesign Criteria 20 and IEEE 279 paragraph 4.1.
The analysis describes instrumentation that is not a part of the SLCS. Justify the non-ccmpliance of the SLCS with the autcmatic actuaticn requirement.
- 2) General Cesign Criteria 28.
The analysis presented dces not address GCC 28. Eces the SLCS meet CCC 28 assuming that the maximun amount of the SLCS piping that could contain cold water dces so at the time the system is activated with the reactor at full power?
3)
IEEE 279, paragraph 4.8 and 4 9 The analysis presented describes instruments that are neither system inputs nor system input senscrs.
C.
RHR.
The analysis refers to the
" analysis for Regulatcry Guide 1.47" for clarificaticn and amplification. What specific Regulatcry Guide 1.47 analysis provides this clarification?
- 2) Regulatory Guide 1 32. The analysis consists of a reference to confermance statements for GCC 17 and IEEE 3C8, neither of which is discussed for RHR.
- 3) Regulatcry Guide 175. What Class 1E RHR cir-uits are redundant to ncn-Class 1E RHR circuits (paragraph e)?
D.
Remote Shutdown System
- 1) Criterion 22, 23, 24, & 29 Identify the " diversity" referred to in these analyses.
I
- 2) TEEE 279, paragraph 4.7, 4.11, 4.12, 4.13, 4.14, 4.16, &
4.19. What is the basis for clatning that these sections are not applicable?
QO31.72 Section 7.4.1 3.3.1 states that during the initial phase of (7.4.1 3) cooling the reactor, only a porticn of the RHR heat exchanger capacity is required. What is the basis for this statement and is it related to the normal operation of the RHR? Also, the second sentence in Section 7.4.1 3.3.2 does not clarify the first sentence of the section. What are the "two diverse shutdchn cooling means" referred to in Secticn 7.4.1 3 3 57 g0j1.,73 The discussion presented does not identify the relation between (7. 4.1. 4) the centrols on the remote shutdown panels and those in the centrol room. Imp,lify your description to indicate khether the two control locaticns operate the sar.e actuators and verify that there are no situations in which the position of a switch in the centrol room can compromise the operability of that device frcm the remote shutdown panel.
e e
e t
L i
t Qo30.74 The tabulation of SRDI in Table 7.5-1 is inconsistent with the (7.5.1) instrumentation identified in the 7.5.1 discussion as being part (T7.1-3) of the SRDI as follows:
(T7.5-1) 1)
7.5.1.2.1 identifies the RPS annunciators as one of the t
(QO31.05) means of verifying shutdown and references Table 7.5-1, which does not include these annunciators.
2) 7.5.1.2.2.1 identifies the fuel zone water level as one of the reactor coolant system redtodant SRDIs, but this intrumentation is anitted from Table 7.5-1.
- 3) Vost of the other 7.5.1.2 discussions identify annunciators, which are not listed in Table 7.5-1, along with the other displays and indicators as providing verification of proper systcm operation.
4) 7.5.1.3 discusses bypassed and inoperable status indication for ESF systems and states that this indication is provided by a combination of lights and annunciators.
The response to Q031.05 indicates that the description of 7.5.13 applies to all instrument systems in Table 7.1-3 for which RG 1.47 is identified as applicable. None of these indicator lights or annunciators ares included in Table 7 5-1.
Amend the FSAR as necessary to provide a consistent and complete identification of the instrumentation that comprises the SRDI.
Q030~75 Resolve the inconsistencies in qualification of SRDI in the (7 5.2.1) various analysis subsections.
The analysis makes the following n
(
(7.5.2.4.2) statements on qualification:
(7.5.2.4.4)
- 1) Section 7.5.2.1 states that " Insofar as practical, the (7.5.2.5.1) subject section instruments are selected from those (7.5.2.5.6) types which are qualifiable."
- 2) Section 7.5.2.4.2 appears to state that all the SRDI, except for safe shutdown instruments, is qualified for operability following a seismic event.
(In most cases the wording is "will function", and in only one para-graph is the wording " qualified to be cperable".)
- 3) Section 7.5.2.4.4 states that for NSSS-engineered safety features, only the isolation valve status is qualified.
- 4) Section 7.5.2.5.1 states that safe shutdown instruments are not fully qualified in the first two paragraphs while paragraph (d) states that all SRDI is qualified, including post-seismic performance.
- 5) Section 7.5.2.5.6 states that the SRDI are the same type and subject to the same qualifications and quality control as the safety systems instruments kr.end your FSAR to correctly identify the SRDI that is qualified, and whether it is qualified seismically as well as environmentally. In addition, unless the safe shutdown instrumentation is the only SRDI that is not seismically qualified, anend your FSAR to:
- 1) Identify the seismic qualification status of each SRDI device when.it is presented in an analysis of SRDI canpliance to satisfy post-seismic accident monitoring
i:
req 4.rements.
- 2) Describe the nethods used to identify the qualified SRDI to the o;.erator and the restrictions imposed on the use of information from unqualified SRDI.
go30.76 The FSAR claims compliance with the single failure criteria of (7.5.2.5.1)
IEEE 279 Since only simple redundancy is provided in the indication of reactor water level and reactor pressure and since these values will change during and following an accident, what instructions or techniques are provided the operator to assure that for any discrepancy in indication the correct signal is identified and used as the basis for operator action (or inaction, as appropriate)?
Qo30.77 Your response to Question 031.05 is inadequate. Contrary to your (7.1.2)
, response, Section 7.1.2.6.9 does not state ecmpliance with (7.2.2)
Regulatory Guide 1.47, it merely refers to Section 7.5.13 for a (7 3 3) description of the control circuits and indicators provided to (7.5.1 3) nenitor bypassed and inoperable status of safety systems.
(7 5.2.2)
Similarly, each analysis for compliance in Sections 7.2, 7.3, and 7.4 only reference Section 7.5.13 Since all of the preceding sections refer to 7.5, and the SRDI described in Section 7.5.1.3 are the means by which Regulatory Guide 1.47 is satisfied, it appears that Section 7.5.2 is the appropriate place to analyze the systems (for which RG 1.47 is applicable as indicated in Table 7.1-3) for compliance. We concur that the guide is not applicable to the other SRDI.
Provide an analysis for canpliance with
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i Regulatory Guide 1.47 in Section 7.5.2.
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